The safety principles applicable to heavy-water, gas-cooled reactors are outlined, with a view to establishing containment specifications adapted to the sites available in Switzerland for the construction of nuclear plants. These specifications are derived from dose rates considered acceptable, in the event of a serious reactor accident, for persons living near the plant, and are based on-meteorological and demographic conditions representative of the majority of the country's sites. The authors consider various designs for the containment shell, taking into account the conditions which would exist in the shell after the maximum credible accident. The following types of shell are studied: pre-stressed concrete; pre-stressed concrete with steel dome; pre-stressed concrete with inner, leakproof steel lining; steel with concrete side shield to protect against radiation; double shell. The degree of leak proofing of the shells studied is regarded as a feature of the particular design and not as a fixed constructional specification. The authors assess the leak proofing properties of each type of shell and establish building costs for each of them on the basis of precise plans, with the collaboration of various specialized firms. They estimate the effectiveness of the various shells from a safety standpoint, in relation to different emergency procedures, in particular release into the atmosphere through appropriate filters and decontamination of the air within the shell by recycling through batteries of filters. The paper contains a very detailed comparison of about 10 cases corresponding to various combinations of design and emergency procedure; the comparison was made using a computer programme specially established for the purpose. The results are compared with those for a reactor of the same type and power, but assembled together with the heat exchangers in a pre-stressed concrete shell. (author) [French] Les principes de securite des reacteurs a eau lourde refroidis

In a nuclear reactor including the bars or plates cooled by a gaseous fluid, burst risks to occur in the sheath assuring the tightness separation between the cooling gas and the fissile materials. It is necessary to be able to detect the formation of these cracks as possible in order to avoid all risk of fission products release or any reaction of uranium to the contact of the refrigerating gas. It is however the increase of the radioactivity in the cooling gas due to the scattering of the fission products that permits to signal the apparition of a crack or to follow its evolution. It is possible to detect cracks of the order of the square millimeter. In this report, we will detail the principle and the realization of a device used for the surveillance of a natural uranium reactor cooled by air circulation. (M.B.) [French] Dans un reacteur nucleaire comportant des barres ou des plaques refroidies par un fluide gazeux des fissures risquent de se produire dans les gaines assurant la separation etanche entre le gaz de refroidissement et les materiaux fissiles. II est necessaire de pouvoir detecter la formation de ces fissures des que possible afin d'eviter tout risque de liberation de produits de fission ou de reaction de l'uranium au contact du gaz refrigerant. C'est cependant l'augmentation de la radioactivite du gaz de refroidissement due a la dispersion des produits de fission qui permet de signaler l'apparition d'une fissure ou de suivre son evolution. On peut ainsi detecter des fissures de l'ordre du millimetre carre. Dans ce rapport, nous detaillerons le principe et la realisation d'un appareil utilise pour la surveillance d'un reacteur a uranium naturel refroidi par circulation d'air. (M.B.)

Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [French] L'eclatement d'un tube de force dans la cuve d'un reacteur de puissance modere a l'eau lourde et refroidi par un gaz sous pression est un accident qui a ete etudie experimentalement a propos d'EL-4. Un premier essai a l'echelle 1 ayant montre que l'eclatement d'un tube ne provoque pas celui des tubes voisins, des essais relatifs a la tenue de la cuve ont ete effectues sur maquettes a echelle tres reduite (l/lO). Il a ete trouve que la cuve peut supporter plusieurs eclatements de tubes sans deformations notables. La pression transitoire dans la cuve a une allure oscillatoire amortie dont le maximum (pression de pic) a fait l'objet d'une etude experimentale detaillee. Il apparait que les parametres essentiels influant sur cette pression sont: la pression du gaz contenu dans le tube de force, le volume du gaz qui participe a l'eclatement, la flexibilite de la cuve, la masse d'air empoisonnee dans la cuve, la nature du gaz explosant. Une methode generale d'estimation des pics de pression dans

In connection with fuel element development work for the high-temperature gas-coolcd reactor of the Brown-Boveri/Krupp Reaktorbau G.m.b.H., two different fuel element concepts were considered and developed. In both cases the fuel element consists of a graphite ball of 6 cm in diam. which contains the fuel insert, a cylindrical pellet of about 20 mm in diam. and 16 mm in height. The two concepts differ in the type of the.fuel insert as well as in the preparation of the graphite ball. In the first concept the fuel insert consists of a mixture of UC{sub 2} and graphite which is prepared by blending U{sub 3}O{sub 8} and graphite, pressing them into pellets and reacting the two components in a vacuum furnace at 1800{sup o}C. The atomic ratio of U : C is 1:45. Since this type of fuel pellet does not retain the fission products completely the surrounding graphite sphere had to be made impervious to fission products by impregnation in order to obtain a fission-product retaining element. Permeabilities of the order of 10{sup -6}cm{sup 2}/s could be achieved. In the second concept the fuel insert consists of a solid solution of UC in ZrC and is coated with a layer of ZrC. The molar ratio of UC to ZrC is 1 : 20. The fuel pellet preparation was accomplished by the following procedure: UO{sub 2}, ZrO{sub 2}, and graphite were mixed and pressed into pellets. The pellets were reacted to the carbides. Ball milling of the carbides was followed by hot pressing at temperatures o f 2000{sup o}C. Densities of more than 95% of the theoretical density could be achieved. A full description of the preparation and of some physical properties of the fuel pellets is given in the paper. A sufficient fission gas retention behaviour of this type of fuel insert which allows it to be put into unimpregnated graphite balls is expected. Other advantages of this kind of fuel are discussed. (author) [French] Dans le cadre des etudes de combustibles destines au reacteur a haute temperature, refroidi par

An important research programme on heavy water reactor physics has been carried out in France for quite a few years. The decision to build the EL 4 prototype and so to choose the heavy water gas cooled type has renewed the interest in this programme and at the same time given to it a more specific orientation A summary of the results gained in this field is presented in this paper. In the first part are described the experimental investigations, most of them were carried out in the criticality facility AQUILON II. The experiments are grouped in four parts - Systematic studies of lattices Buckling measurements. - Specific studies of gas-cooled lattices. - Fine structure, spectral indices measurements etc... - Measurements on lattices or samples containing Uranium of various enrichment or Plutonium. The second part is devoted to a summary of the theoretical studies. The whole results have allowed an improvement of the calculation methods, have led to a better understanding of the neutron balance in lattices, and have permitted the establishment of a set of formula to predict not only the clean fuel conditions but also the evolution of the nuclear properties with irradiation. Some specific studies on power reactor are quoted. (authors) [French] Un important programme d'etudes sur la physique des reacteurs a eau lourde est mene en France depuis assez longtemps. La decision de construire le prototype EL 4 et de s'engager ainsi dans la filiere des reacteurs a eau lourde refroidis par gaz a redonne un nouvel interet a ce programme et l'a en meme temps oriente dans une direction plus particuliere. La presente communication, rassemble les resultats des etudes faites dans ce domaine depuis la derniere conference de Geneve. Dans la premiere partie on decrit les etudes experimentales dont la plupart ont ete effectuees dans la pile d'experiences critiques Aquilon II. Les experiences sont groupees en quatre ensembles: etude systematique de reseaux (mesures de laplaciens) etudes

The present thesis covers an applied study on tire pyrolysis. The main objective is to develop tools to allow predicting the production and the quality of oil from tire pyrolysis. The first research objective consisted in modelling the kinetics of tires pyrolysis in a reactor, namely an industrial rotary drum operating in batch mode. A literature review performed later demonstrated that almost all kinetics models developed to represent tire pyrolysis could not represent the actual industrial process with enough accuracy. Among the families of kinetics models for pyrolysis, three have been identified: models with one single global reaction, models with multiple combined parallel reactions, and models with multiple parallel and series reactions. It was observed that these models show limitations. In the models with one single global reaction and with multiple parallels reactions, the production of each individual pyrolytic product cannot be predicted, but only for combined volatiles. Morevoer, the mass term in the kinetics refers to the final char weight (Winfinity) that varies with pyrolysis conditions, which yields less robust models. Also, despite the fact that models with multiple parallels and series reactions can predict the rate of production for each pyrolysis product, the selectivities are determined for operating temperatures instead of real mass temperatures, giving models for which parameters tuning is not adequate when used at the industrial scale. A new kinetics model has been developed, allowing predicting the rate of production of noncondensable gas, oil, and char from tire pyrolysis. The novelty of this model is the consideration of intrinsic selectivities for each product as a function of temperature. This hypothesis has been assumed valid considering that in the industrial pyrolysis process, pyrolysis kinetics is limiting. The developed model considers individual kinetics for each of the three pyrolytic products proportional to the global decomposition kinetics of pyrolysables. The simulation with data obtained in industrial operation showed the robustness of the model to predict with accuracy in transient regime, tires pyrolysis, with the help of model parameters obtained at laboratory scale, namely in regards of the trigger of production, the residence time of tires (dynamic production) and the amount of oil produced (cumulative yield). It is a novel way to model pyrolysis that could be extrapolated to new waste materials. The second objective of this doctoral research was to determine the evolution of specific tires specific heat during pyrolysis and the enthalpy of pyrolysis. The origin of this objective comes from a primary contradiction. With few exceptions, it is acknowledged that organic materials pyrolysis is globally an endothermic phenomenon. At the opposite, all experiments led with laboratory apparatuses such as DSC (Differential Scanning Calorimetry) showed exothermic peaks during dynamic experiments (constant heating rate). It has been confirmed by results obtained at the industrial scale, where no sign of exothermicity has been observed. The Hess Law has also confirmed these results, that globally, pyrolysis is indeed a completely endothermic process. An accurate energy balance is required to predict mass temperature during pyrolysis, this parameter being unbindable from kinetics. An advanced investigation of char first allowed demonstrating that specific heat of solids during pyrolysis decreases with increasing temperature until the weight loss peak is reached, around 400°C, and then starts increasing again. This observation, combined with the fact that the sample loses weight during pyrolysis is considered as the major cause of the apparition of an exothermic peak in laboratory scale experiments. That is, the control system of these apparatuses generates a bias and an unavoidable overheat of the samples producing this exothermic behavior. It would thus be an artifact. On the base of new data on the evolution of global specific heat during pyrolysis, a model of the energy balance has been developed at the industrial scale to determine the enthalpy of pyrolysis. The simulation has shown that a major part of the heat transferred to the pyrolized mass would make its temperature increase. Next, an enthalpy of pyrolysis dependent of weight loss was obtained. Finally, two other terms of enthalpy have been found, namely an enthalpy for the breakage of sulfur bridges and an enthalpy for the stabilization of char when conversion approaches completion. This research will have allowed establishing a novel general methodology to determine the enthalpy of pyrolysis. More particularly, new clarifications hasve been obtained in regards to the evolution of specific heat of solids during pyrolysis and new enthalpies of pyrolysis, all endothermic, could be obtained, in agreement with the theoretical expectations. The third research objective concerned the behavior of sulfur during tires pyrolysis. With as a premise that sulfur is an intrinsic contaminant of many types of waste, it is critical to clarify its fate during pyrolysis, in the present case for waste tires. It has been observed in the literature that some quantitative analyses had been presented, but generally, the mechanisms for the distribution of sulfur within the pyrolytic products remain unclear. Thus, it was then not possible to predict the transfer of sulfur to each of the tire pyrolysis products. The results taken form literature have been complemented with a series of TGA experiments followed by complete elemental analyses of the residual solids. Mass balances have been performed in order to characterize the distribution of elements within the three products (noncondensable gas, oil, and char). A novel parameter has been created during this research: the sulfur loss selectivity. This intrinsic selectivity is a prediction of the distribution of sulfur within the pyrolysis products as a function of temperature. Three phenomena has been identified that could affect the sulfur loss selectivity. First, the natural devolatilization of sulfur due to pyrolysis. Next, the sulfur devolatilization due to the desulfurization of the solid matrix by hydrogen and finally, the clustering of sulfur in the solid state due to metal sulfidation (zinc and iron). The results have shown that this selectivity reach a limit value of 1 when pyrolysis is limited by the kinetics and in the absence of metal. When the mass transfer is limiting at low temperature (<500°C) the selectivity will be greater than 1. At a temperature over 350°C with the presence of metals, the selectivity will be lower than 1. It is a useful tool for industrial pyrolysis processes, being a novel indicator for the distribution of contaminants during the pyrolysis of waste. A better comprehension of these mechanisms allows elaborating a better strategy when designing these industrial processes. For example, in light of this research, it could be preferable to pre-treat the tires at lower temperature to eliminate a significant part of sulfur before pyrolyzing them at high temperature. The resulting pyrolytic products would then necessitate a lighter purification post-treatment, being more efficient and more economical.

In three articles attention is paid to the use of synthesis gas in the Netherlands. In the first article the growing market for synthesis gas is discussed, in the second article it is argued that the energy efficiency of the production of nitrogen can be improved, and the third article is about cooling an ice track by means of carbon dioxide. [Dutch] In 3 artikelen wordt aandacht besteed aan het gebruik van industriele gassen. In het eerste artikel wordt de sterk groeiende markt voor synthese gas besproken, in het tweede artikel wordt aangegeven hoe de efficiency van stikstofproduktie kan worden verbeterd, en het derde artikele gaat in op het gebruik van kooldioxide in de koeling van een ijsbaan.

'Aquilon' is an experimental reactor specially designed for the neutronic study of heterogeneous multiplying media with solid fuel and liquid moderator. Since this study is in general incompatible with energy production, the power of the reactor has been limited to a minimum so as to be able to obtain a simple and compact structure, easy access, good handling and great flexibility of operation and utilisation. (author) [French] 'Aquilon' est un reacteur experimental specialement concu pour l'etude neutronique de milieux multiplicateurs heterogenes a combustible solide et ralentisseur liquide. Cette etude etant en general incompatible avec la production d'energie, on a limite au minimum la puissance du reacteur pour pouvoir obtenir une structure simple et peu encombrante, un acces facile, une bonne maniabilite et une grande souplesse de fonctionnement et d'utilisation. (auteur)

In industrial buildings a balanced ventilation is not common property. Due to increasing differences in pressure across the front much cold outside air enters the building, resulting in poor comfort near the door. A good air curtain can heat up large amounts of incoming cold outside air in such a way that energy saving is optimal and comfort remains high. (mk) [Dutch] Bij industriele gebouwen is gebalanceerde ventilatie geen gemeengoed. Door oplopende drukverschillen over de gevel komt veel koude buitenlucht naar binnen en is het comfort vlak achter de deur erg laag. Een goed luchtgordijn kan grote hoeveelheden binnenkomende koude buitenlucht zodanig opwarmen dat de energiebesparing optimaal is en het comfort hoog blijft.

In a heavy-water moderated reactor cooled by pressurized gas, an important problem from the point of view, of the reactor block and its economics is the choice of the gas supply system. In the pressure tube solution, the whole of the reactor block structure is at a relatively low temperature, whereas the gas supply equipment is at that of the gas, which is much higher. These parts, through which passes the heat carrying fluid have to present as low a resistance as possible to it so as to avoid costly extra blowing power. Finally, they may only be placed in the reactor block after it has been built; the time required for putting them in position should therefore not be too long. The work reported here concerns the various problems arising in the case of each channel being supplied individually by a tube at the entry and the exit which is connected to a main circuit made up of large size collectors. This individual tubing is sufficiently flexible to absorb the differential expansion and the movement of its ends without stresses or prohibitive reactions being produced; the tubing is also of relatively short length so as to reduce the pressure head of the pressurized gas outside the channels; the small amount of space taken up by the tubing makes it possible to assemble it in a manner which is satisfactory from the point of view both of the time required and of the technical quality. (authors) [French] Dans un reacteur modere a l'eau lourde et refroidi au gaz sous pression, un probleme important du point de vue du trace du bloc pile et de son economie est le choix du systeme d'alimentation en gaz. Pour une solution a tubes de force, l'ensemble des structures du bloc reacteur est a temperature relativement faible, alors que les organes d'alimentation en gaz sont a celle, notablement plus elevee, du gaz. Ces organes, traverses par le debit du caloporteur, doivent lui opposer le minimum de resistance afin de ne pas necessiter un supplement onereux de puissance de

A study is made of the general case of the theory of first-order isotopic chemical exchange between a gaseous and a liquid phase in a reactor, starting from fundamental reaction kinetics data, and without making any limiting hypothesis concerning the value of the separation factor. The cases of counter-current reactors and of co-current reactors are considered successively. The general deuterium conservation equation requires the definition of the quotient of the reactor; the performances of this reactor are characterised by its overall efficiency. The idea of the ratio is introduced because it represents a convenient intermediary in the calculations. The search for an additive value for reactors in series leads logically to the defining of an exchange capacity, and a total efficiency, or number of theoretical reactors. This method of expressing the performances of a reactor is more general than the efficiency due to Murphee which only has a physical significance in the particular case of homogeneous liquid reactors. The relationships between these various quantities are established, and the representation due to Mc Cabe and Thiele is generalized. The reactor performances are linked to the first - order reaction kinetics by the transfer number. The relationships are given for a certain number of concrete cases. Finally the application of these calculations is given, together with the approximations necessary in the case where, because of the presence of several components in each phase, the exchange reaction no longer obeys a single kinetic law. (authors) [French] On examine dans le cas general la theorie d'un reacteur quelconque pour l'echange chimique isotopique du premier ordre entre une phase gazeuse et une phase liquide, a partir des donnees fondamentales sur la cinetique de la reaction, sans faire aucune hypothese limitative sur le cas des reacteurs a contre ourant, puis celui des reacteurs a co-courant. L'equation generale de conservation du deuterium

Two types of chemical treatment can be considered for the manufacture of solid sources for industrial uses from fission product solutions remaining after plutonium extraction: a) concentration of the solution and preparation of solid sources from the bulk material, without separation, b) separation of one or several fission products from which the sources are made. Examination of the radio-chemical composition of the mixture of fission products that will be available from the Marcoule reactors (G1, G2 and G3) shows that caesium-137 accounts for 30 per cent of the {gamma} energy available immediately after the plutonium separation, 70 per cent two years after and 100 per cent after five years. There is practically no advantage in making sources from bulk fission products, since the separation of caesium-137 is no more complicated and yet it results in a material with more potential uses. The separation of caesium-137 by a method based on the standard phospho-tungstate precipitation method has been considered. Previously, the precipitated caesium phospho-tungstate was dissolved and caesium was recovered from the solution by cation-exchange or by removal of phosphate and tungstate ions by anion-exchange. A study has now been made, of the metathesis of caesium phospho-tungstate to barium phosphate and tungstate by the action of barium hydroxide, the caesium being obtained in solution as the hydroxide. The advantages of this new procedure are: - greater decontamination of caesium-137 without further purification, - possibility of direct transformation to caesium sulphate, - general simplification of the procedure and, consequently, of the equipment. (author) [French] Deux types de traitement chimique peuvent etre envisages pour amener a l'etat de sources solides utilisables industriellement les produits de fission contenus dans les solutions residuaires de l'extraction du plutonium. Ces traitements sont les suivants: a) concentration des solutions et confection de

Two types of chemical treatment can be considered for the manufacture of solid sources for industrial uses from fission product solutions remaining after plutonium extraction: a) concentration of the solution and preparation of solid sources from the bulk material, without separation, b) separation of one or several fission products from which the sources are made. Examination of the radio-chemical composition of the mixture of fission products that will be available from the Marcoule reactors (G1, G2 and G3) shows that caesium-137 accounts for 30 per cent of the {gamma} energy available immediately after the plutonium separation, 70 per cent two years after and 100 per cent after five years. There is practically no advantage in making sources from bulk fission products, since the separation of caesium-137 is no more complicated and yet it results in a material with more potential uses. The separation of caesium-137 by a method based on the standard phospho-tungstate precipitation method has been considered. Previously, the precipitated caesium phospho-tungstate was dissolved and caesium was recovered from the solution by cation-exchange or by removal of phosphate and tungstate ions by anion-exchange. A study has now been made, of the metathesis of caesium phospho-tungstate to barium phosphate and tungstate by the action of barium hydroxide, the caesium being obtained in solution as the hydroxide. The advantages of this new procedure are: - greater decontamination of caesium-137 without further purification, - possibility of direct transformation to caesium sulphate, - general simplification of the procedure and, consequently, of the equipment. (author) [French] Deux types de traitement chimique peuvent etre envisages pour amener a l'etat de sources solides utilisables industriellement les produits de fission contenus dans les solutions residuaires de l'extraction du plutonium. Ces traitements sont les suivants: a) concentration des solutions et

The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [French] Le Commissariat a l'Energie Atomique (dans le cadre du plan quinquennal) a entre autres objectifs, la realisation des deux premiers reacteurs francais moderes au graphite. La construction du reacteur G-1 a Marcoule, premiere pile plutonigene francaise, est realise afin qu'il puisse diverger au debut de 1956 et atteindre sa pleine puissance au debut du second semestre de la meme annee. Dans ce rapport nous detaillerons les specificites du reacteur et en particulier son systeme de refroidissement et de recuperation d'energie. Le reacteur G-1 etant essentielement destine a permettre aux techniciens francais d'etudier le plus tot possible le comportement d'une installation productrice d'energie empruntant sa chaleur a une source nucleaire. (M.B.)

We studied the variations of power and reactivity of a reactor when we raise in a continuous way the starting plates. During the subcritical regime (negative reactivity), the power is determined by reactivity and by the intensity of the sources of photo neutrons, produced during the previous work of the reactor. When, during the rise of the plates, the reactor, pass by the critical regime (zero reactivity), one notes that the reached power is independent of the initial reactivity. During the sur-critical regime (positive reactivity), the elevation of temperature of the uranium bars slows down the growth of reactivity due to the movements of the plates. The power stretches then toward a value that depends only on the regime of cooling of the reactor and the excess of the available reactivity. This survey permits to choose such a rise speed, that reactivity remains constantly lower to a value beyond which the piloting of the reactor becomes difficult. This result is not more valid, if the intensity of the sources is insufficient, what takes place during the first divergences and after a stop of long length. (author) [French] On etudie les variations de puissance et de reactivite d'un reacteur quand on leve d'une facon continue les plaques de demarrage. Pendant le regime subcritique (reactivite negative), la puissance est determinee par la reactivite et par l'intensite des sources de photoneutrons, produites pendant la marche anterieure du reacteur. Quand, au cours de la montee des plaques, le reacteur passe par le regime critique (reactivite nulle), on constate que la puissance atteinte est independante de la reactivite initiale. Pendant le regime surcritique (reactivite positive), l'elevation de temperature des barres d'uranium ralentit l'accroissement de reactivite due aux mouvements des plaques. La puissance tend alors vers une valeur qui ne depend plus que du regime de refroidissement du reacteur et de l'exces de la reactivite disponible. Cette etude permet de

The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [French] La puissance d'un reacteur nucleaire, dans les conditions du regime, est affectee de fluctuations dont les causes sont tres diverses. Ce comportement aleatoire rentre dans le cadre general de l'etude des 'bruits'. Entre autres sources ce bruit, nous analysons ici les fluctuations dues: a) a l'emission discontinue des neutrons provenant d'une source autonome; b) a la multiplication des neutrons au sein du reacteur. La methode que nous introduisons exploite les analogies entre les lois qui regissent un reacteur nucleaire au regime et certains systemes radioelectriques, en particulier les circuits a boucle de reaction. Le reacteur est caracterise par sa 'bande passante' et est decrit comme un systeme soumis a une succession d'impulsions aleatoires. Dans les conditions de fonctionnement non lineaires, l'effet du bruit neutronique est precise en utilisant une fonctionnelle non lineaire, ce qui relie cette theorie a

The thermal efficiency of subject systems, especially at higher temperatures is influenced by heat and humidity transport underground. Thermal conductivity and specific thermal capacity depend on the humidity content of the soil. A simulation model was developed that describes the coupled heat and humidity transport in the temperature range up to 90 C. This model will be validated in laboratory and field tests and then be used for designing and analysing underground stores. Pilot plants for the storage of industrial waste heat were designed and planned on the basis of this simulation. In both cases these are cogeneration plants whose waste heat was to be used for space heating and as process energy. Both plants have a very high demand of electric energy which is mostly supplied by the cogeneration plant. The waste heat is put into the store during the summer. In the winter heat is supplied by both the store and the cogeneration plant. In both cases the store has a volume of approx. 15,000 cubic metres with 140 and 210 pits located in a depth of 30 and 40 metres. The plants are used to carry out extensive measurements for the validation of simulation models. (orig.) [Deutsch] Die thermische Leistungsfaehigkeit solcher Systeme wird insbesondere im hoeheren Temperaturbereich durch den Waerme- und Feuchtetransport im Untergrund beeinflusst. Sowohl die Waermeleitfaehigkeit als auch die spezifische Waermekapazitaet sind vom Feuchtegehalt des Bodens abhaengig. Es wurde ein Simulationsmodell entwickelt, das den gekoppelten Waerme- und Feuchtetransport im Temperaturbereich bis 90 C beschreibt. Dieses Modell wird an Labor- und Feldexperimenten validiert und dient dann zur Auslegung und Analyse von Erdwaermesonden-Speichern. Basierend auf diesen theoretischen Grundlagenarbeiten wurden Pilotanlagen zur saisonalen Speicherung industrieller Abwaerme ausgelegt und geplant. In beiden Faellen handelt es sich um Kraft/Waermekopplungsanlagen, deren Abwaerme zur Gebaeudeheizung und

An attempt has been made to develop certain ideas concerning homogeneous reactors. The project under consideration is based on the simultaneous use of a suspension of uranium dispersed in heavy or light water and of boiling in the reactor for heat extraction. However, the studies of suspensions and of boiling are relatively independent and can also be developed for reactors of different types using one or the other. Our aim is a minimum investment in fissile material; for this we propose to extract the steam directly from the core and to make use of a cyclone to accelerate this extraction; a cyclone-type circulation creating a field of increasing tangential velocities of the fluid towards the axis causes the droplets of vapour to accelerate towards the axial vortex in which they are collected; the steam output is then evacuated to the external heat utilisation system, for example an exchanger of the condenser-boiler type. The input speed of water into the reactor being one of the important parameters in the running of the pile, a spiral supply input chamber is used, allowing this speed to be regulated in amount and direction. (author)Fren. [French] Nous nous sommes attaches a developper certaines idees relatives aux piles homogenes. Le projet que nous etudions est base sur l'emploi simultane d'une suspension contenant de l'uranium disperse dans l'eau legere ou lourde et de l'ebullition dans le reacteur pour l'extraction de chaleur. Neanmoins, les etudes de suspensions et d'ebullition sont relativement independantes et peuvent egalement etre developpees pour des reacteurs de type different utilisant l'une ou l'autre. Le but que nous cherchons a atteindre est un investissement minimum en matiere fissile; pour cela, nous proposons d'extraire directement la vapeur dans le coeur et de recourir a un dispositif cyclone pour accelerer cette extraction; une circulation type cyclone creant un champ de vitesses tangentielles du fluide croissantes veraxe a pour effet d

Full Text Available La sociologie du travail a développé un point de vue particulier sur l'analyse des temporalités sociales. L'objectif de cet article est de montrer l'élaboration progressive d'une critique d'un temps industriel unique et unifiant. Cette critique s'appuie sur trois éléments : la multiplicité des temporalités de l’activité professionnelle, le caractère déstructurant du travail industriel et l'émergence d'un temps des loisirs. Ce point de vue a été élaboré par la sociologie du travail en France après 1945, notamment par Georges Friedmann, Pierre Naville et William Grossin. Avant d’aborder cette période nous voudrions interroger le moment de la fondation de la sociologie. En mobilisant des travaux peu connus de Max Weber sur le travail, nous verrons en quoi il est un précurseur de l’analyse des attitudes diversifiées des individus et des collectifs à l’égard du temps. Cette perspective lie, dès les premières enquêtes sociologiques, le travail industriel à une interrogation sur les temporalités.The sociology of work and labor developed a particular viewpoint on the analysis of social times. The objective of this article is to show how a criticism of a unique and unifying industrial time progressively grew up. Such criticism leans on three elements: the multiplicity of social times linked to the occupational activity, the disintegrating nature of work in the manufacturing process and the emergence of leisure-time. This point of view was elaborated by the sociology of work in France after 1945, notably by Georges Friedmann, Pierre Naville and William Grossin. Before approaching this period we question the moment of the foundation of sociology. By mobilizing little known research by Max Weber on the manufacturing process, we shall see in what way he was a precursor of the analysis of diversified attitudes regarding time by individuals and groups.

A study is made of the problem of the flow-rate and pressure distributions along the length of two volumes, internal and external, bounded by a series of non-continuous annular elements placed along the channel axis. It is observed that the phenomenon can easily be represented by equations. The theoretical expressions observed are particularly simple when the distances between the elements are above a certain minimum value. The experimental work has made it possible to show that the theoretical formulation derived is valid with a very great accuracy. The experimental study has also been carried out in the case of a very small spacing between the elements. It has been possible to show in this case that the hypothesis made for deriving the theoretical expressions was perfectly justified. In the last part finally, we consider the practical problem of evaluating the pressure-drops between the ends of a series of annular elements. (author) [French] On etudie le probleme de la repartition des debits et des pressions le long des deux espaces, interne et externe, delimites par une succession d'elements annulaires non jointifs, disposes suivant l'axe d'un canal. On constate que le phenomene peut etre mis aisement en equation. Les relations theoriques obtenues prennent en particulier une forme simple lorsque les intervalles entre elements sont superieurs a une valeur minimum. L'etude experimentale a permis de constater que cette formulation theorique etablie etait alors valable avec une precision excellente. L'etude experimentale a ete egalement effectuee dans le cas de tres faibles intervalles entre elements consecutifs. On a pu alors verifier que l'hypothese adoptee dans l'etablissement des relations theoriques etait parfaitement justifiee. Dans un dernier chapitre enfin, nous abordons le point de vue pratique de l'evaluation des pertes de charge aux bornes d'un train d'elements annulaires. (auteur)

A study is made of the problem of the flow-rate and pressure distributions along the length of two volumes, internal and external, bounded by a series of non-continuous annular elements placed along the channel axis. It is observed that the phenomenon can easily be represented by equations. The theoretical expressions observed are particularly simple when the distances between the elements are above a certain minimum value. The experimental work has made it possible to show that the theoretical formulation derived is valid with a very great accuracy. The experimental study has also been carried out in the case of a very small spacing between the elements. It has been possible to show in this case that the hypothesis made for deriving the theoretical expressions was perfectly justified. In the last part finally, we consider the practical problem of evaluating the pressure-drops between the ends of a series of annular elements. (author) [French] On etudie le probleme de la repartition des debits et des pressions le long des deux espaces, interne et externe, delimites par une succession d'elements annulaires non jointifs, disposes suivant l'axe d'un canal. On constate que le phenomene peut etre mis aisement en equation. Les relations theoriques obtenues prennent en particulier une forme simple lorsque les intervalles entre elements sont superieurs a une valeur minimum. L'etude experimentale a permis de constater que cette formulation theorique etablie etait alors valable avec une precision excellente. L'etude experimentale a ete egalement effectuee dans le cas de tres faibles intervalles entre elements consecutifs. On a pu alors verifier que l'hypothese adoptee dans l'etablissement des relations theoriques etait parfaitement justifiee. Dans un dernier chapitre enfin, nous abordons le point de vue pratique de l'evaluation des pertes de charge aux bornes d'un train d'elements annulaires. (auteur)

The various methods used during the physical study of the reactor G1 to determine the variations of the effective multiplication factor consecutive to a given change in the geometry of the multiplying medium, are presented and discussed. The comparison of the results obtained by these various methods has allowed their validity to be tested and precise conditions of use to be given. In the first part are presented the principles used and their ranges of validity. In the second part the experimental results are given, together with some indications on their comparison with theoretical estimations. (author) [French] Nous exposons et discutons diverses methodes utilisees, lors de l'etude physique du reacteur G1, pour determiner les variations du facteur de multiplication effectif consecutives a un changement donne dans la geometrie du milieu multiplicateur. La comparaison des resultats obtenus par diverses methodes nous a permis de tester leur validite et d'en preciser les conditions d'emploi. Dans une premiere partie, nous exposons les principes utilises et leurs domaines de validite. Dans une seconde partie nous donnons les resultats experimentaux obtenus avec quelques indications sur leur comparaison avec les estimations theoriques. (auteur)

The solution to the correlation equations are expressed in terms of the eigenvalues and Eigen-matrices of the transport operator, for a subcritical zero power reactor. This allows to define, for each point of the reactor and for detectors detecting neutrons of given velocities, correlation and transfer functions driven by the same white-noise source. A precise meaning is also given to the importance operator, which is the adjoin of the transport operator. (author) [French] La solution des equations regissant les matrices de correlation est exprimee en fonction des valeurs et matrices propres de l'operateur de transport pour un reacteur sous-critique et de puissance nulle. Ceci permet de definir, en chaque point du reacteur et pour des detecteurs repondant a des neutrons de vitesse definie, des fonctions de correlation et de transfert dont les entrees sont attaquees par une meme source de bruit blanc. Le role joue par l'operateur importance, adjoint de l'operateur de transport, est aussi precise. (auteur)

The introduction, at an increasing rate, of power reactors using slightly enriched uranium will inevitably lead to the production of considerable quantities of plutonium over the next decade. Fast reactors will not be capable of absorbing this material before 1980. The question thus arises of whether one should store the plutonium far future use in fast reactors, recycle it in existing thermal reactors, or try to sell it. The problem has been studied for an electric power generating system that does not foresee selling the plutonium produced by its reactors and does not buy plutonium outside, which enables a good approximation to be made and eliminates the major unknown quantity represented by the future market price of plutonium. Assuming within this system a programme that provides for the construction of power reactors of a given type and capacity at specific dates, the utilization of the plutonium produced can be optimized by linear programming techniques so as to minimize the discounted total cost of the power generated over a given period. A later stage consists in optimizing, by various techniques, not only the utilization but also the production of plutonium by appropriate selection of the power reactor types to be constructed. (author) [French] L'implantation, a un rythme croissant, de centrales nucleaires a uranium legerement enrichi entrainera la production ineluctable d'une quantite importante de plutonium au cours de la prochaine decennie. Les reacteurs a neutrons rapides ne seront capables d'absorber cette production qu'apres 1980. La question se pose donc de savoir s'il est preferable de stocker le plutonium en vue de son utilisation ulterieure dans les reacteurs a neutrons rapides plutot que de le recycler dans les reacteurs actuels a neutrons thermiques ou d'essayer de le vendre. Ce probleme a ete etudie dans le cadre d'un systeme de production d'energie electrique qui ne prevoirait pas la vente du plutonium produit par ses reacteurs nucleaires ni

After demonstrating the sensitivity of the distribution of power in a production reactor to a deformation caused by dissymmetries of reactivity in the reactor, the authors describe the method of neutron flux determination devised for the reactors G2 and G3 under working conditions; the detector used is a tungsten or nickel wire, the {gamma} activity of which is measured with an ionisation chamber. Several flux determinations are given as examples to illustrate the sensitivity of the method. (author) [French] Apres avoir mis en evidence la sensibilite de la repartition de la puissance dans un reacteur de production a une deformation provoquee par de faibles dissymetries de reactivite dans le reacteur, les auteurs decrivent la methode de releve du flux neutronique mise au point pour les reacteurs G2 et G3 en puissance; le detecteur utilise est un fil de tungstene ou de nickel dont l'activite {gamma} est mesuree a l'aide d'une chambre d'ionisation. Quelques releves de flux illustrant la sensibilite de la methode sont donnes a titre d'exemple. (auteur)

This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios ({approx}1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 {delta}k/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [French] Ce rapport presente les etudes qu'il nous a paru important d'aborder dans le cadre du developpement des reacteurs a neutrons rapides. - Il met en evidence l'interet des taux de regeneration internes eleves ({approx}1, 1) pour obtenir une bonne evolution dans le temps de la distribution de puissance et de la reactivite (moins de 0,005 {delta}k/k pour 18 mois). - Il montre la possibilite d'y parvenir tout en applatissant la distribution des fissions, ce qui se traduit par une puissance specifique moyenne plus elevee (gain economique), et donc un temps de doublement plus faible de l'ordte de 10 ans - Il tente de definir un optimum de la puissance specifique valable pour les

This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios ({approx}1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 {delta}k/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [French] Ce rapport presente les etudes qu'il nous a paru important d'aborder dans le cadre du developpement des reacteurs a neutrons rapides. - Il met en evidence l'interet des taux de regeneration internes eleves ({approx}1, 1) pour obtenir une bonne evolution dans le temps de la distribution de puissance et de la reactivite (moins de 0,005 {delta}k/k pour 18 mois). - Il montre la possibilite d'y parvenir tout en applatissant la distribution des fissions, ce qui se traduit par une puissance specifique moyenne plus elevee (gain economique), et donc un temps de doublement plus faible de l'ordte de 10 ans - Il tente de definir un optimum de la puissance specifique valable pour les projets de reacteurs futurs

with hierachical social relations between industrialists and workers, specifying a set of duties and obligations for each of the two parties to meet. Industrialists were, among other things, expected to care for workers who were no longer able to provide for themselves due to illness or old age. The article...... discusses different schemes for taking care of ill and aged workers and shows that such schemes were relatively common around 1900. The discussion focuses on norms and practices, not effects. Whether the schemes actually resulted in lessened social tensions during the transformation to an industrial society...

Melusine II, low-power reactor, used for the study of Siloe core has diverged at the CEA Grenoble, the 23. May 1962; its monitoring board studied and carried out in this center is the first in France to be entirely transistorized. The first months of running have justified the hope put in the new electronics to improve the stability and the safety of running. The article describes the design of the control and gives the main characteristics of the measurement chains and of the actions on reactivity. (O.M.) [French] Melusine II, reacteur de faible puissance destine a l'etude du coeur de Siloe a diverge au Centre d'Etudes Nucleaires de Grenoble, le 23 mai 1962, son tableau de controle etudie et realise dans ce Centre est le premier en France a etre entierement transistorise. Les premiers mois de fonctionnement ont justifie l'espoir mis dans la nouvelle electronique pour ameliorer la stabilite et la surete de fonctionnement. L'article decrit la conception du controle et donne les principales caracteristiques des chaines de mesure et des actions sur la reactivite. (auteurs)

A pulsed fast reactor (IBR) has been operating at rated capacity since December 1960 in the Joint Institute for Nuclear Research. This reactor is used as a pulsed neutron source for physical experiments carried out by the time-of-flight method. It is used for total cross-section and intermediate neutron capture cross- section measurements, for studying the interaction between slow neutrons and solids and liquids, and for measuring neutron spectra produced in various media. The paper describes the basic structural features of the reactor and the results of the experiments for which it has been used. The reactor's operating system is based on recurrent pulses. Power pulses are produced when the mobile part of the reactor core moves swiftly through the stationary part of the core. The mobile part of the core is fastened to a rotating disc and travels at a speed of 230 m/s. The frequency of power pulses can be altered by means of an auxiliary mobile zone which has a range of 2.3-88 pulses per second. The mean power of the reactor is 1 kW, and the half-width of the power pulse in 36 {mu}s. The reactor is provided with a control and safety system which ensures automatic maintenance of mean power and swift shutdown in the event of any operational irregularity. It is fitted with a system of evacuated-neutron-flight tubes used in time-of-flight experiments. The main tube is 1000 m in length. In the start-up process and during physical experiments carried out on the reactor, the influence on reactivity of displacing the controls and the mobile parts of the core was studied ; the length of the pulse was measured under various operating conditions, and power pulse amplitude fluctuations were studied. Further measurements were made to establish the lifetime of prompt neutrons, the effective fraction of delayed neutrons, and coefficients of reactivity. (author) [French] L'Institut unifie de recherches nucleaires dispose d'un reacteur puise a neutrons rapides (IBR), qui

This report examines the various fuel handling systems which presently seem feasible for a fast power reactor. It tries to point out the advantages and / or the the disadvantages and the fabrication problems for each solution involved and makes, a tentative to evaluate the time required for a fuel loading and / or unloading operation. One has investigated the influence of the maximum allowable irradiation, the number of of shut-downs, the power distribution shape within the core on the storage capacity needed, the load factor expected and the average irradiation obtained. (authors) [French] On a examine dans ce rapport les differents systemes de manutention, qui semblent actuellement realisables pour un reacteur a neutrons rapides de puissance, en essayant de faire ressortir les avantages, les inconvenients et les difficultes de realisation de chaque systeme, et de chiffer les temps de manutention auxquels ils conduisent. On a aussi regarde l'influence des variations du taux d'irradiation maximal,de la cadence des arrets ou de la forme du flux dans le coeur du reacteur, sur la capacite du stockage, le taux de disponibilite et le taux d'irradiation moyen. (auteurs)

This report examines the various fuel handling systems which presently seem feasible for a fast power reactor. It tries to point out the advantages and / or the the disadvantages and the fabrication problems for each solution involved and makes, a tentative to evaluate the time required for a fuel loading and / or unloading operation. One has investigated the influence of the maximum allowable irradiation, the number of of shut-downs, the power distribution shape within the core on the storage capacity needed, the load factor expected and the average irradiation obtained. (authors) [French] On a examine dans ce rapport les differents systemes de manutention, qui semblent actuellement realisables pour un reacteur a neutrons rapides de puissance, en essayant de faire ressortir les avantages, les inconvenients et les difficultes de realisation de chaque systeme, et de chiffer les temps de manutention auxquels ils conduisent. On a aussi regarde l'influence des variations du taux d'irradiation maximal,de la cadence des arrets ou de la forme du flux dans le coeur du reacteur, sur la capacite du stockage, le taux de disponibilite et le taux d'irradiation moyen. (auteurs)

The {alpha} decay constant of prompt neutrons has been studied in the homogeneous plutonium-fueled, light-water-moderated reactor Alecto, by the probability method. In this method, the probability to count one, two,.... neutrons during a given time is measured. The value of {alpha} can be deduced from this measurement, for various subcritical states of the reactor. The experimental results were then compared with values obtained, for the same reactivities, by the pulsed neutron technique. (authors) [French] On a etudie sur Alecto, reacteur homogene au plutonium, modere a l'eau legere, la constante de decroissance {alpha} des neutrons prompts par la methode des probabilites. Celle-ci consiste a mesurer la probabilite de compter un, deux, etc..., neutrons pendant un intervalle de temps donne. On a pu en deduire la valeur de {alpha}, dans divers etats sous-critiques du reacteur. On a compare les resultats experimentaux a d'autres valeurs obtenues, aux memes reactivites, par la methode des neutrons pulses. (auteurs)

This work is related to the CEA `Escrime` project which concerns the reliability and functioning safety of nuclear reactors, and in particular the operation and supervision of nuclear installations. Its aim is the analysis and the formalizing of PWRs operation in order to define the collaboration and optimum sharing of tasks between human operators and automatized systems for an improved functioning safety. Chapter 1 describes the operation of nuclear reactors and the instrumentation and control activities. It focusses on the weaknesses of actual automatized systems and examines the interest of the multi-agents approach to build an improved automatized system. Chapter 2 presents the actual state of the art about multi-agent systems and about their application to reactor operation. Chapter 3 is devoted to the definition of the conceptual model of automatized systems developed in this work (distribution of operation activities, competition between agents, hierarchy, arbitration). Chapter 4 describes the computer model of the essential operating system elaborated according to the conceptual model defined above. Modeling is performed using Spirit and an application is described in chapter 5. (J.S.). 58 refs.

The chemical effect of ionizing radiations on a certain number of gaseous systems is described. Under the influence of radiations from a reactor, NH{sub 3}, is decomposed to nitrogen and hydrogen in stoichiometric proportions. Formation of N{sub 2}H{sub 3}, particularly could not be detected. Under a slow neutron flux the reaction {sup 14}N (n, p) {sup 14}C constitutes the main source of decomposition energy. Direct recombination of H, and N, has been brought about under the influence of radiation. The radiolysis of NH{sub 3}, occurs by a complex mechanism; and the kinetics follow a law of the order of about 2.5 which increases with the decomposition rate. The decomposition of hydrogen sulphide is appreciably faster than that of NH{sub 3}. Hydrogen is the only gaseous product of the reaction. The sulphur, which is deposited on the walls of the ampoules, is clearly visible to the naked eye. Up to the present decompositions up to 84 per cent have been obtained. The influence of the reaction {sup 32}S (n, p) {sup 32}P is considered. Radiochemical decomposition of nitrous oxide N{sub 2}O takes place with high yields. The reaction is complicated from the beginning by the formation of higher oxides of nitrogen which we identify and measure. Radiochemical decomposition of methane gives quantities of higher hydrocarbons. Certain of these gaseous systems could find applications in the measurement of high doses of radiation. This problem is discussed in the conclusion. (author)Fren. [French] L'effet chimique des rayonnements ionisants sur un certain nombre de systemes gazeux est decrit. Sous l'influence des rayonnements d'un reacteur, NH{sub 3} se decompose en azote et hydrogene en proportions stoechiometriques. En particulier aucune formation de N{sub 2}H{sub 4}, n'a pu etre detectee. Sous flux de neutrons lents, la reaction {sup 14}N (n, p){sup 14}C constitue la principale source d'energie de decomposition. La recombinaison directe de H{sub 2} et N{sub 2} a ete realisous l

In this paper, a new formulation of the spatial dependent impulse response of a subcritical reactor in a cylindrical geometry is proposed. An expression of the transfer function between a point source at the center of coordinates and the flux at a given point (r,z) is obtained by solving: by means of Laplace transform, the one group diffusion equation. In this transfer function, variables r and p (p being the Laplace variable) remain linked within a modified Bessel function. Taking the inverse Laplace transform is done by two different ways: - using the Mellin-Fourier method which separates variables r and t. This method makes it possible to establish that there is identity between the classical formulation and the new one. - using an inverse Laplace transform which keeps variables r and t linked. This method requires to approximate the inverse Laplace transform of the end factor. It is then possible to replace the radial harmonics modes series of the classical expression by a single function. This new formulation seems to be of particular interest when dealing with reactors of large size and lifetime. It is also interesting each time the harmonics play an important role. (author) [French] Dans le present rapport, on propose une nouvelle formulation de la reponse impulsionnelle spatio-temporelle d'un reacteur sous-critique, en geometrie cylindrique. Une expression de la fonction de transfert entre une source ponctuelle placee au centre des coordonnees et le flux au point courant (r,z) est obtenue en resolvant, par transformation de Laplace, l'equation de la diffusion a un seul groupe d'energie. Dans cette fonction de transfert, les variables r et p (variable de Laplace) demeurent groupees dans une fonction de Bessel modifiee. Le retour a l'original est effectue de deux manieres: - la methode de Mellin-Fourier qui separe les variables r et t, permet d'etablir l'identite entre la nouvelle formulation et la formulation classique. - un original conservant les variables

We have previously developed a method for studying neutronic fluctuations in nuclear reactors using the analogy between the behaviour of a reactor and that of certain common radioelectric circuits. The fluctuations may then be calculated by introducing into the circuit a suitable noise source. By this method we have been able to consider the overall fluctuations in a particularly simple form and we have provided a physical significance for certain results obtained more laboriously by other methods. The object of the present report is to generalise this method and in particular to extend it to the case of a reactor having a cellular structure and to apply it to fluctuations within a cell. It is thus shown that the fluctuations in a cell are the resultant of two terms: - a rapidly evolving Poissonian noise, not related to the overall fluctuations; - a slowly evolving noise, when the reactor is not too far from criticality, which is related to the overall fluctuations. The first term arises from a rapid 'ordering' of the system, during which time the cells come mutually into equilibrium. The second term is due to the coordinated evolution of all the cells, after the end of the first transitory phase. The conclusions reached show that it would be useful to complete the study with an analysis of non-linear phenomena which can considerably influence the transitory behaviour of the cells during the initial pre-equilibrium phase. This report also Stresses the relationship of the new method to the old methods. It tends also to place pile fluctuation theory in a more general framework, that of the fluctuations of a system depending on several random parameters; from this point of view, the method could easily be transposed and adapted to the study of other physical problems of this type. (authors) [French] Nous avons precedemment developpe une methode d'etude des fluctuations neutroniques des reacteurs nucleaires mettant a profit l'analogie entre le comportement d

From all the methods possible for controlling thin cans the one chosen, for numerous reasons, vas that making use of ultrasonic techniques. A method has been developed which should make it possible to carry out a rapid and efficient industrial control of canning tubes, The reasons for the choice of the ultrasonic method are given in detail, together with the principles of the method and the actual control parameters. In the present state of our research, it should be possible to control at least 50 000 tubes a year. Improvements brought about in the details of the control technique itself should make it possible to increase this rate considerably. (authors) [French] Parmi toutes les methodes possibles de controle des gaines minces, le procede retenu pour de multiples raisons a ete celui faisant appel a la technique des ultrasons. Une methode a ete mise au point qui doit permettre un controle industriel rapide et efficace des tubes de gaine. Sont exposes en detail, les raisons du choix de la methode par ultrasons, les principes de cette methode et les parametres du controle proprement dit. Dans l'etat actuel de nos etudes la cadence devrait permettre le controle de 50000 tubes par an au minimum. Des ameliorations de detail portant sur la technique de controle elle-meme, doivent permettre d'accelerer tres notablement cette cadence. (auteurs)

In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors) [French] Dans le volume 1: Ce rapport est une description du reacteur Pegase, afin d'examiner la surete des installations. Le site de CADARACHE ou elles sont situees, a ete sommairement decrit, en particulier, a cause des consequences sur les techniques mises en oeuvre pour la realisation de Pegase. Nous nous sommes egalement attache a decrire les aspects originaux du reacteur. Les boucles autonomes destinees a tester en vraie grandeur des elements combustibles de la filiere uranium naturel graphite-gaz, ainsi que leurs dispositifs de controle et d'exploitation, figurent egalement dans ce rapport. Dans le volume 2: Dans le present rapport, nous examinons des accidents pouvant endommager des installations du reacteur Pegase. Les origines d'accidents examines

In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors) [French] Dans le volume 1: Ce rapport est une description du reacteur Pegase, afin d'examiner la surete des installations. Le site de CADARACHE ou elles sont situees, a ete sommairement decrit, en particulier, a cause des consequences sur les techniques mises en oeuvre pour la realisation de Pegase. Nous nous sommes egalement attache a decrire les aspects originaux du reacteur. Les boucles autonomes destinees a tester en vraie grandeur des elements combustibles de la filiere uranium naturel graphite-gaz, ainsi que leurs dispositifs de controle et d'exploitation, figurent egalement dans ce rapport. Dans le volume 2: Dans le present rapport, nous examinons des accidents pouvant endommager des installations du reacteur Pegase. Les origines d'accidents examines comprennent la seismicite

The eventual application of the technique of cryogenic coils for the production of magnetic fields for liquid hydrogen bubble chambers has led us to the study and construction of two copper coils (cooled in a bath of liquid hydrogen). Part one treats certain generalities concerning the resistivity of several pure metals at very low temperatures, the gain in power, thermal isolation and the choice of conductor and power supply. Part two describes the experimental arrangement. The last part is devoted to a description of the two coils which were constructed, and to the results obtained. (author) [French] L'application eventuelle de la technique des ''bobines cryogeniques'' aux chambres a bulles a hydrogene liquide pour la production du champ magnetique nous a conduits a l'etude et a la construction de deux bobines en cuivre (refroidies dans un bain d'hydrogene liquide). La premiere partie traite de generalites concernant la resistivite de quelques metaux purs, a tres basse temperature, le gain de puissance, l'isolement thermique et le choix du conducteur et de l'alimentation electrique. Dans la seconde partie on decrit l'installation d'essais. La derniere partie est consacree a la description des deux bobines realisees et aux resultats obtenus. (auteur)

The comparative study of the "apprentissage industriel" in French speaking Belgium and of the "contrat de qualification jeunes" in France shows the maintenance of the neo corporatist model of social State. The analysis of the emergence and of the implementation of these two employment public policies for low qualified young permit to understand how they are appropriated on the ground, especially in the metallurgy sector. If these policies are intended for those low qualified young, the implem...

1) Introduction: The difficulties of the formulation of the equations of phenomena occurring during the operation of a fusion reactor are underlined. 2) The possibilities presented by analog computation of the solution of nonlinear differential equations are enumerated. The accuracy and limitations of this method are discussed. 3) The analog solution in the stationary problem of the measurement of the discharge confinement is given and comparison with experimental results. 4) The analog solution of the dynamic problem of the evolution of the discharge current in a simple case is given and it is compared with experimental data. 5) The analog solution of the motion of an isolated ion in the electromagnetic field is given. A spatial field simulator used for this problem (bidimensional problem) is described. 6) The analog solution of the preceding problem for a tridimensional case for particular geometrical configurations using simultaneously 2 field simulators is given. 7) A method of computation derived from Monte Carlo method for the study of dynamic of plasma is described. 8) Conclusion: the essential differences between the analog computation of fission reactors and fusion reactors are analysed. In particular the theory of control of a fusion reactor as described by SCHULTZ is discussed and the results of linearized formulations are compared with those of nonlinear simulation. (author)Fren. [French] 1) Introduction. On souligne les difficultes que presente la mise en equation des phenomenes mis en jeu lors du fonctionnement d'un reacteur a fusion. On selectionne un certain nombre d'equations generalement utilisees et on montre les impossibilites analytiques auxquelles on se heurte alors. 2) On rappelle les possibilites du calcul analogique pour la resolution des systemes differentiels non lineaires et on indique la precision de la methode ainsi que ses limitations. 3) On decrit esolution analogique du probleme statique de la mesure du confinement de la decharge

The main objective of this study is to publicise the data obtained during five years operation of the reactor G2 and G3 at Marcoule as far as the cooling gas is concerned, from storage of reserves up to its slow escape into the atmosphere, and including all the stages of its practical use, its chemical examination, its nuclear behaviour and its possible physicochemical transformation. This work can not only yield information about the operations carried out at Marcoule but can also provide useful suggestions for improving the sealing and for decreasing the activity of the pressurized gas circuits in reactors similar to G2/G3. (authors) [French] Le but principal de cette etude est de diffuser les connaissances acquises au cours de cinq annees d'exploitation des reacteurs G2 et G3 de Marcoule en ce qui concerne le gaz de refroidissement, depuis son stockage d'appoint jusqu'a son echappement lent dans l'atmosphere, en passant par tous les stades de son utilisation pratique, de son etude chimique, de son comportement nucleaire, eventuellement de ses transformations physico-chimiques. Cette etude peut, non seulement renseigner sur les operations effectuees couramment a Marcoule, mais egalement donner des suggestions interessantes pour l'amelioration de l'etancheite et la diminution de l'activite des circuits de gaz en pression dans des reacteurs analogues a G2/G3. (auteurs)

The power limit of swimming-pool type reactors depends on the phenomenon of the appearance of burn-out. In order to determine this limit we have attempted to detect the local boiling which usually occurs before the burn out. Local boiling has been simulated by an electrically heated plate placed in the core of the reactor Siloette. The study of local boiling, which is based on the properties of the correlation functions for the neutron noise of detectors placed in the core, shows that a privileged frequency occurs in the power spectrum of the noise. It is intended in the future to determine the influence of various parameters on this characteristic frequency. (author) [French] La limitation de la puissance des reacteurs nucleaires de type piscine est due au phenomene d'apparition de 'burn out'. Pour determiner cette limitation, nous nous sommes proposes dans ce rapport de detecter l'ebullition locale qui apparait generalement avant le 'burn out'. L'ebullition locale a ete simulee par une plaque chauffee electriquement et placee dans le coeur du reacteur SILOETTE. L'etude de l'ebullition locale, qui est basee sur les proprietes des fonctions de correlation du bruit neutronique de detecteurs places clans le coeur, fait apparaitre une frequence privilegiee dans le spectre de puissance du bruit. On envisage dans l'avenir, de determiner l'influence des divers parametres sur cette frequence caracteristique. (auteur)

Corrosion of aluminium in contact with moderating water in nuclear reactor leads to the formation of an alumina hydrosol which can have an adverse effect on the operation of the reactor. Several physical methods have been used in an attempt to counteract this effect. The method proposed here consists in the elimination of the aluminium by dissolution and subsequent fixation in the ionic form on mixed-bed ion-exchange resin. In order to do this, the parameters and the values of these parameters most favorable to the dissolution process have been determined. If the moderator is heavy water, the deuterated acid can be prepared by converting a solution in heavy water to a salt of the acid using a deuterated cationic resin. (author) [French] La corrosion de l'aluminium au contact de l'eau moderatrice des reacteurs nucleaires, donne lieu a la formation d'un hydrosol d'alumine nuisible au bon fonctionnement des reacteurs. Plusieurs methodes physiques ont ete mises en oeuvre pour pallier ces inconvenients. On propose ici d'eliminer l'alumine par solubilisation pour la fixer ensuite sous forme ionique par des resines echangeuses d'ions, en lit melange. A cette fin on determine les parametres et leurs grandeurs favorables a cette solubilisation. Si le moderateur est de l'eau lourde la preparation d'acide deutere peut etre effectuee par passage d'une solution en eau lourde a un sel de l'acide sur resine cationique deuteree.

After resuming briefly the characteristics of the installations G2-G3 at Marcoule and EDF 1 at Chinon, the authors review the main aspects of the tests, the starting and the exploitation of these reactors. Among the various points examined, particular emphasis is given to the devices of original nature such as tubular fuel elements, flattening of the neutron flux by stuffing, behaviour of the reactor tanks and the cooling circuits, the blowers, unloading devices, regulation and functioning of the informations. This analysis deals equally with the performances obtained and the difficulties and the various incidents experienced during the initial starting period. Among the more interesting results, the progressive increase in the power of the Marcoule reactors is mentioned, obtained through a better knowledge of the parameters covering the functioning of the reactors such as the distribution of the flux and the temperatures etc... acquired during the course of the exploitation of the reactor. The conclusion reached by the authors is that the experience gained on these installations has shown: - that during an initial period, adjustments became necessary, all of which turned out to be possible, - that an analysis of their functioning has permitted the progressive movement towards a truly industrial exploitation. (authors) [French] Les auteurs, apres un bref rappel des caracteristiques des installations G2 - G3 de MARCOULE et E.D.F. 1 de CHINON, passent en revue les principaux aspects des essais, de la mise en service et de l'exploitation de ces centrales. Parmi les divers points examines, une attention speciale est accordee aux dispositifs presentant un caractere original tels que elements combustibles tubulaires, aplatissement du flux neutronique par gavage, comportement des caissons des reacteurs et des circuits de refroidissement, soufflantes, appareils de dechargement, regulation et fonctionnement des informations. L'analyse presentee porte tant sur les

1) The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2) Starting from this concept, we endeavoured to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3) Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author) [French] 1) La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2) A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3) Enfin une methode de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)

1. The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2. Starting from this concept, we endeavored to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3. Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author)Fren. [French] 1. La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2. A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3. Enfin une mde de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)

This apparatus is designed to measure instantaneously the positive or negative reactivity of a uranium reactor moderated by light water, on condition that the point of departure is the critical state of the reactor, or an already known sub-critical state. Slight modifications only are required to adapt it to another type of reactor. It is an analogue computer which simply inverses the transfer function of the reactor; it is not therefore a model reactor of which the output voltage is connected by a servo-mechanism to the power of the reactor to give the reactivity; the principle of the calculation of the reactivity does not depend on a servomechanism. One of its disadvantages is that it cannot operate outside a power variation range of 2.5 decades. However the measurement of a negative reactivity value between 0 and 3000 pcm is immediate. It measures the reactivity without deducting it from the period; it therefore gives the reactivity very precisely both for divergence and convergence even through in this latter case the period does not in fact exist. The equipment makes it possible to calibrate very rapidly the control rods of a reactor (the rod-drop method), to measure the reactivity of an experiment in the core, and to measure certain temperature effects. It is also possible by introducing a control into the core at a measured rate, to deduce directly its efficiency curve. (author) [French] Cet appareil est destine a mesurer instantanement la reactivite positive ou negative d'un reacteur a uranium modere a l'eau legere, a condition de partir de l'etat critique du reacteur, ou eventuellement d'un etat sous-critique deja connu. De legeres modifications permettent de l'adapter a un autre type de moderateur. C'est un calculateur analogique, qui inverse purement et simplement la fonction de transfert du reacteur; ce n'est donc pas un simulateur de pile dont la tension de sortie est asservie a la puissance du reacteur pour elaborer la reactivite; le principe du

organic liquids under irradiation. The equipments of Melusine are now being modernized. Siloe, another swimming-pool reactor, has been operating at 15 MW since the end of 1963. The performances achieved constitute a considerable progress in the field of swimming-pool reactors, since the fluxes obtained with Siloe are in the same order of magnitude than those which needed till now a tank type structure, whose numerous disadvantages in the fitting of experiments are well known. Siloe will be used mostly in the study of structural materials, graphites, refractory fissile materials, and for solid state physics. The reactor Pegase, in service in the Cadarache Nuclear Centre since 1963, is intended solely for testing full-scale fuel elements of EDF and EL 4 types. The current programme, for the eight loops of the reactor, covers the elements for the reactors EDF 2, EDF 3 and EL 4. New loops are in the course of being studied for the fuel elements of the EDF 4 and EDF 5 reactors. The general line of the CEA programmes has shown up the considerable need for fast neutron irradiations. The reactor Osiris which is in the course of being constructed will serve to complement the CEA's equipment in this field and at the same time fill the gap which will be left in the near future in the Centre de Saclay by the closing down of EL 2. Osiris is a light water reactor whose special structure will allow it to function at 50 MW without the disadvantages usually associated with the presence of a heavy waterproof tank. This reactor, which should be put into service in 1966, is mainly intended for the investigation of structural materials, graphite and refractory fuels; it will also serve to increase the production of high specific activity isotopes, and to develop activation analysis techniques. (authors) [French] Les auteurs examinent successivement les differents reacteurs de recherche en service dans les Centres du Commissariat a l'Energie Atomique. Ils retracent brievement l'histoire de ces

The detector described is designed for the instantaneous measurement of thermal neutron fluxes, in the presence of high {gamma} ray activity; this detector can withstand temperatures as high as 500 deg. C. It is based on the following principle: radioactive atoms resulting from heavy-nucleus fission are carried by a gas flow to a detector recording their {beta} and {gamma} disintegration. Thermal neutron fluxes as low as few neutrons per cm{sup 2} per second can be measured. This detector may be used to control a nuclear reactor, to plot the thermal flux distribution with an excellent definition (1 mm{sup 2}) for fluxes higher than 10{sup 8} n/cm{sup 2}/s. The time response of the system to a sharp variation of flux is limited, in case of large fluxes, to the transit time of the gas flow between the fission product emitter and the detector; of the order of one tenth of a sec per meter of piping. The detector may also be applied for spectroscopy of fission products eider than 0,1 s. (author)Fren. [French] On decrit un appareil permettant la mesure instantanee des flux de neutrons thermiques accompagnes de flux intenses de rayons {gamma} et situes dans des enceintes pouvant etre portees a des temperatures superieures a 500 deg. C. On utilise la radioactivite des atomes resultant de la fission des noyaux lourds; ces atomes sont entraines par un courant gazeux vers un detecteur de radioactivite qui enregistre leurs desintegrations {beta} et {gamma}. On peut mesurer des flux partir de quelques neutrons thermiques par cm{sup 2} et par seconde. L'appareil permet de suivre la puissance d'un reacteur atomique, de tracer des cartes de densite de neutrons avec une tres bonne definition (1 mm{sup 2}) dans le cas de flux superieurs a 10{sup 8} cm{sup 2}/s. Le temps de reponse du systeme a une variation du flux de neutrons est limite, poes flux importants, par le temps de transit du gaz entre l'emetteur de produits de fission et le detecteur: soit quelques dizaines de

The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors) [French] L'experience acquise par l'exploitation des reacteurs de MARCOULE, la construction et le demarrage des reacteurs d

Experience of the CEA heavy water reactors and a systematic study of the radiolytic decomposition of water in the core of swimming-pool reactors are described. Setting up of reactivity control by physico-chemical methods. Reactivity control by homogeneous poisoning of the reactor A comparison of the evolution of xenon poisoning with the residual anti reactivity of the poison in solution during its nuclear consumption establishes the programme which must govern the variation in its concentration if the exact compensation is to be produced The behaviour of the poison towards the reactor materials under the particular operational conditions must be taken into account. Radiolytic decomposition of water in the reactors in the presence of soluble poisons: A study of the effect of certain chemically inert salts, present in small concentrations in the water, on its radiolytic decomposition rate, has led to some new results which are discussed. The choice of a soluble poison is justified on the basis of the above results. Reactivity control by the use of a gaseous absorbent The use of a gas control rod circuit for compensation purposes, in place of solid control rods is described. The use of soluble poisons in the moderator to compensate the xenon effect, and of a gaseous absorbent in a circuit known as a gas control rod form original aspects of the reactivity control in the reactor EL 4. (authors) [French] L'observation du comportement de l'eau et de l'eau lourde dans les reacteurs en exploitation, contribue au fonctionnement sur de ceux-ci et oriente certaines etudes relatives aux techniques de controle de la reactivite par mise en oeuvre de poisons solubles. L'utilisation de poisons nucleaires dissous dans l'eau du reacteur entraine une pollution chimique de celle-ci. Les conditions d'emploi permettant d'eviter les effets indesirables de cette pollution sont etudiees. Les problemes analytiques - bien qu'importants - ne sont pas abordes dans le cadre de la communication

The subjects is an experimental out-of pile loop study of the hydrodynamic oscillations occurring in boiling-water reactors. The study was carried out at atmospheric pressure and at pressure of about 8 atmospheres, in channels heated electrically by a constant and uniform specified current. In the test at 8 atmospheres the channel was a round tube of approximately 6 mm interior diameter. At 1 atmosphere a ring-section channel was used, 10 * 20 mm in diameter, with an inner heating tube and an outer tube of pyrex. It was possible to operate with natural convection and also with forced convection with test-channel by-pass. The study consists of 3 parts: 1. Preliminary determination of the laws governing pressure-drop during boiling. 2. Determination of the fronts at which oscillation appears, within a wide range of the parameters involved. 3. A descriptive study of the oscillations and measurement of the periods. The report gives the oscillation fronts with natural and forced convection for various values of the singular pressure drop at the channel inlet and for various riser lengths. The results are presented in non-dimensional form, which is available, in first approximation, for all geometric scales and for all fluids. Besides the following points were observed: - the wall (nature and thickness) can be an important factor ; - oscillation can occur in a horizontal channel. (author) [French] II a ete effectue une etude experimentale, en boucle hors-pile, des oscillations hydrodynamiques survenant dans les reacteurs a ebullition. L'etude a ete effectuee a la pression atmospherique et a une pression voisine de 8 atmospheres dans des canaux chauffes electriquement a puissance imposee constante et uniforme. Dans les essais a 8 atmospheres le canal etait un tube circulaire de diametre interieur 6 mm environ. A 1 atmosphere le canal etait de section annulaire 10 * 20 mm avec un tube interieur chauffant et un tube exterieur en pyrex. Le fonctionnement etait possible

The subjects is an experimental out-of pile loop study of the hydrodynamic oscillations occurring in boiling-water reactors. The study was carried out at atmospheric pressure and at pressure of about 8 atmospheres, in channels heated electrically by a constant and uniform specified current. In the test at 8 atmospheres the channel was a round tube of approximately 6 mm interior diameter. At 1 atmosphere a ring-section channel was used, 10 * 20 mm in diameter, with an inner heating tube and an outer tube of pyrex. It was possible to operate with natural convection and also with forced convection with test-channel by-pass. The study consists of 3 parts: 1. Preliminary determination of the laws governing pressure-drop during boiling. 2. Determination of the fronts at which oscillation appears, within a wide range of the parameters involved. 3. A descriptive study of the oscillations and measurement of the periods. The report gives the oscillation fronts with natural and forced convection for various values of the singular pressure drop at the channel inlet and for various riser lengths. The results are presented in non-dimensional form, which is available, in first approximation, for all geometric scales and for all fluids. Besides the following points were observed: - the wall (nature and thickness) can be an important factor ; - oscillation can occur in a horizontal channel. (author) [French] II a ete effectue une etude experimentale, en boucle hors-pile, des oscillations hydrodynamiques survenant dans les reacteurs a ebullition. L'etude a ete effectuee a la pression atmospherique et a une pression voisine de 8 atmospheres dans des canaux chauffes electriquement a puissance imposee constante et uniforme. Dans les essais a 8 atmospheres le canal etait un tube circulaire de diametre interieur 6 mm environ. A 1 atmosphere le canal etait de section annulaire 10 * 20 mm avec un tube interieur chauffant et un tube exterieur en pyrex. Le fonctionnement etait possible en

The design of a two-halves zero-energy fast reactor is briefly described, particular emphasis being placed on those features which determine the practicability and precision of reactor physics measurements. The advantages and disadvantages of the design are discussed with reference to the two years' operating experience of the reactor. The following topics are dealt with: the experimental convenience of the lay-out and of the two halves design; the size and precision of the fuel pieces and the accuracy of location of the fuel elements; the effects of edge irregularities and heterogeneity of structure on the accuracy with which the critical mass of an 'ideal' equivalent assembly is determined; reproducibility of the critical condition after dismantling the assembly, or separating the two halves; variation of reactivity with separation of the halves, including effects of asymmetric loading; sensitivity of various counters, neutron source strength, use of an accelerator neutron source; speed of response of safety circuits and consequent restrictions on rate of assembly of the two halves; additional precautions necessary in using plutonium fuel; and notes on the accuracy of measurement of reactivity and on the practical limitations affecting various other reactor physics measurements. (author) [French] Les auteurs decrivent brievement ce modele de reacteur a neutrons rapides et de puissance zero construit en deux moities, en insistant particulierment sur les caracteristiques qui determinent la possibilites de faire des mesures relatives a la physique des reacteurs et la precision de ces mesures. Ils exposent les avantages et les inconvenients de ce modele compte tenu de l'experience acquise au cours des deux annees de fonctionnement du reacteur. Ils traitent les sujets suivants: interet pratique, au point de vue experimental, du plan de ce reacteur et de sa constitution en deux moities; dimension et precision des pieces de combustible et exactitude de l'emplacement des

The solidification by bitumen of chemical coprecipitation sludges from the Marcoule waste treatment station has been studied in the laboratory and has led to the construction of an industrial coating plant. The quality of the coated material obtained has been controlled by the lixiviation test carried out with ordinary water and with sea-water on 45 ml laboratory samples and on industrial coated blocks of 150 litres. Tests on blocks of such a size have necessitated the installation of three special tanks. Two, each of 2000 litres capacity, contain ordinary and sea-water which was continuously recycled at a rate of 2.5 cm/hr and renewed periodically. In the third tank having a capacity of 11000 litres, the coated block was buried in earth and sprinkled with ordinary water with a view to studying the migration of radioelements in soil. The results of these tests confirm those obtained during the laboratory experiments. (authors) [French] La solidification par le bitume des boues de coprecipitation chimique de la station de traitement des effluents du Centre de Marcoule, etudiee en laboratoire, a conduit a la realisation d'une installation industrielle d'enrobage. La qualite de l'enrobe obtenu a ete controlee par le test de lixiviation qui a ete effectue en eau ordinaire et en eau de mer sur des echantillons de laboratoire de 45 ml et sur des blocs d'enrobe industriel de 150 litres. L'experimentation sur des blocs de telles dimensions a necessite l'installation de 3 cuves speciales. Deux, d'une capacite de 2000 litres, contiennent de l'eau ordinaire et de l'eau de mer recyclees en permanence a une vitesse de 2.5 cm/h et renouvelees periodiquement. Dans la 3eme cuve d'une capacite de 11000 litres, le bloc d'enrobe a ete enfoui dans de la terre et asperge d'eau ordinaire afin d'etudier la migration des radioelements dans le sol. Les resultats de ces essais confirment ceux obtenus au cours des tests de laboratoire. (auteurs)

After a theoretical study on the slowing down of a centrifugal pump, the motion equations have been checked by means of experimental tests. In order to have important slowing down times (which is the case of the cooling pumps of a research reactor) it is necessary to add a fly-wheel. To prevent troubles when starting, a block pump-fly-wheel with clutch under controlled torque was developed. It is so possible to start the fly-wheel progressively without increasing too much power of the driving motor. (author) [French] Apres une etude theorique sur le mouvement de ralentissement d'une pompe centrifuge, les equations du mouvement ont ete verifiees par des essais pratiques. Pour obtenir des temps de ralentissement importants (cas des pompes de refrigeration d'un reacteur de recherche) il est necessaire d'y adjoindre un volant d'inertie. Pour eviter les inconvenients au demarrage, on a etudie un ensemble pompe-volant avec embrayage sous couple controle. Cette solution permet de lancer progressivement le volant sans augmentation appreciable de la puissance du moteur d'entrainement. (auteur)

The apparatus studied is a digital calculating assembly which makes it possible to prepare and to present numerically the period and power of a nuclear reactor during operation, from start-up to nominal power. The pulses from a fission chamber are analyzed continuously, using real time. A small number of elements is required because of the systematic use of a calculation technique comprising the determination of a base 2 logarithm by a linear approximation. The accuracy obtained for the period is of the order of 14%; the response time of the order of the calculated period value. An approximate value of the power (30%) is given at each calculation cycle together with the power thresholds required for the control. (author) [French] L'appareil etudie est un ensemble de calcul digital permettant d'elaborer et d'afficher numeriquement la periode et la puissance, d'un reacteur nucleaire lors de son fonctionnement depuis le demarrage jusqu'a la puissance nominale. Il traite en temps reel, de facon continue, les impulsions en provenance d'une chambre de fission. Grace a l'utilisation systematique d'une technique de calcul, la determination d'un logarithme a base 2 par approximation lineaire, un nombre reduit d'elements est utilise. La precision obtenue sur la periode est de l'ordre de 14 pour cent, le temps de reponse de l'ordre de la valeur de la periode calculee. Un ordre de grandeur de la puissance (30 pour cent) est donne a chaque cycle de calcul ainsi que des seuils de puissance necessaires au controle. (auteur)

Canada's first nuclear power station, the Nuclear Power Demonstration station (NPD), is intended to serve as a means of proof-testing the performance of the Canadian type of station using natural uranium as fuel and heavy water as moderator and coolant. It reached full power on 28 June 1962. Although designed for base-load operation it will, during the early stages, be operated part of the time on high-capacity.runs and part of the time on improvement periods. Progress has been favourable so far; the first high-capacity run of six weeks'duration yielded a capacity factor of 70%. Improvements already made have increased safety, improved performance and demonstrated potential methods of capital-cost reduction for future stations. For example, shaft seals on primary coolant pumps have been modified for better performance, freezer-type vapour recovery equipment has been replaced in favour of absorption columns to reduce heavy-water vapour loss, and flow limiters are being installed in sample lines to reduce losses of heavy water in the event of joint failures. During December 1962 two simultaneous leaks from the on-power refuelling machine led to an unusual sequence of events in which a considerable amount of hot high-pressure heavy water was spilled into the reactor vault where it suffered a slight downgrading in isotopic purity. It was upgraded and the reactor returned to operation by the end of the month. All safety devices operated correctly during the incident as did the provisions for containment of heavy water. (author) [French] La premiere centrale nucleaire du Canada, NPD, est une centrale de demonstration, qui doit servir a verifier les performances des reacteurs fonctionnant a l'uranium naturel et utilisant de l'eau lourde comme ralentisseur et comme fluide de refroidissement. Elle a atteint sa pleine puissance le 28 juin 1962 bien que concue pour etre exploitee comme centrale de base, elle fonctionnera au debut comme centrale d'appoint, ce qui permettra d

Industrial-scale tests are reported of the efficiency of two recovery processes for the separation of uranium from sulfuric acid pickling solutions used on ore from Bois-Noirs, at the Gueugnon works. The final stage of each process is sodium uranate. The earlier part of the report deals with tests of the separation of uranium from foreign metals by fractional precipitation. The second part deals with the separation of uranium from these metals by carbonation of the solutions. (author) [French] Le present rapport concerne les essais industriels de deux procedes de recuperation de l'uranium de solutions d'attaque sulfurique du minerai des Bois-Noirs a l'usine de Gueugnon. Le stade final pour ces deux procedes etant l'uranate de sodium, une premiere partie est consacree aux essais de separation de l'uranium des metaux etrangers par precipitation fractionnee; une deuxieme partie est consacree aux essais de separation de l'uranium des metaux etrangers par carbonatation des solutions d'attaque du minerai. (auteur)

This report is a kind of audit report on a research laboratory, the DER (Departement d'Etudes des Reacteurs, Reactor Study Department) whose activity if focused on four main themes: neutron transport simulation in reactor cores, thermal-hydraulic simulation of reactors, design and safety of innovative reactors, nuclear instrumentation for reactors. The authors discuss an assessment of the whole unit activities in terms of strengths and opportunities, aspects to be improved, risks and recommendations, productions and publications, scientific quality, influence and attractiveness (awards, recruitment capacity, capacity to obtain financing and to tender, participation to international programs), strategy and governance, and project. These same aspects are then discussed and commented for each theme

The isotopic composition of plutonium to be used as fuel for fast reactors will depend on the source of plutonium. In principle three different sources are possible: (a) production reactors; (6) thermal power reactors (using natural uranium or enriched uranium as fuel); (c) fast reactor blankets. In general, source (a) and to some extent source (c) will provide relatively 'clean' plutonium, that is mostly Pu{sup 239}, while plutonium from source (6) will be 'dirty' plutonium, that is plutonium rich in Pu{sup 240}, Pu{sup 241}, and Pu{sup 242}. The degree of 'dirtiness' will depend on the kind of reactor, amount of burn-up and in general on the irradiation history of the fuel. The question then arises, can one use as fuel for fast reactors any kind of plutonium? To investigate the effect of different isotopic composition of the plutonium fuel, in the metallic, oxide and carbide form, on the performance of fast reactors, a limited series of spherical geometry 16-group diffusion theory calculations were performed, using the 16-group cross-section set developed recently by Yiftah, Okrent and Moldauer and taking three different kinds of plutonium, starting with pure Pu{sup 239} and increasing the amount of higher isotopes. For the systems studied-800, 1500 and 2500-l core-volumes, which are typical for large fast power reactors-the result is, when one takes into account only the thermally fissionable isotopes Pu{sup 239} arid Pu{sup 241}, that the 'dirtier' the plutonium, the smaller the critical mass and the higher the breeding ratio. For the 1500-l reactor, taken as an example, it is further found that in the metallic, oxide and carbide plutonium fuels the reactivity change upon removal of 40% of the sodium initially present in the core is made more negative (or less positive) when the plutonium is richer in higher isotopes. (author) [French] La composition isotopique du plutonium qui doit etre utilise comme combustible dans des reacteurs a neutrons rapides depend de

Facing the current weather and energy global problem, Canada chose to develop a reactor cooled by water at supercritical conditions (SCWR). Inspired by the current CANDU-6 pressure tube technology, this concept should allow to save a substantial amount of efforts for developping a brand new kind of reactor by using the well-known pressure tube CANDU design. In fact, this type of reactor should be able to reach a better energy efficiency as well as other essential criteria about safety, security, non-proliferation... Nevertheless, there are still a lot of technology challenges to be dealt with to satisfy the differents obligations related to the use of supercritical water (SCW). Thus, materials to use remain undetermined because of a 25 MPa operating pressure and a 650.C temperature for the SCW coolant. Actually, materials in presence of SCW should be able to avoid too much corrosion and remain low neutrons absorbers. Additionnaly, the use of a light water coolant makes the neutronic absorption more important than in CANDU heavy-water cooled reactors. Additionally, a positive coolant void reactivity (CVR) and safety related problem remains among the challenges to overcome for developping a SCWR. Bringing about a solutions to all these problems remains very difficult and that's why some concessions on these criteria have to be made in order to achieve a viable reactor. This study presents some thougts and works in that direction. Originally developped in early studies about thermodynamic cycle optimization for a SCW power plant, a new hybrid reactor concept with two channels types has arise. To this purpose, we imagine a pressure tube core design but with two different types of channels: . Some channels have thermodynamic conditions where water goes through a supercritical state. . The other channels have "CANDU like" thermodynamic conditions allowing the flow of pressurized light water under sub-critical conditions. These two kinds of features should mitigate the drawbacks and constraints of a whole SCWR core while keeping some of the benefit of the SCW use (improved energy efficiency). The aim of this study is to determine whether or not this kind of hybrid core with two types of channels is viable. To this purpose we used neutronics as thermalhydraulic models. First of all, in this thesis, we describe the SCWR-development history and specific back- ground with bibliography commentaries and a global description of the hybrid reactor developped. Secondly, we provide a brief theorical review of the neutronics and thermalhydraulics concepts used in the DRAGON, DONJON and ARTHUR calculation codes. Afterwards we will be considering some cell and core models and their description. Among others, some coupled neutronics-thermalhydraulic calculations have been done. This is followed by a presentation of the results we obtained and a discussion of the viability of our reactor. Finallly, we conclude.

The 'Reactor Jules Horowitz' is a new research reactor project dedicated to materials and nuclear fuels testing, the location of which is foreseen at the CEA-Cadarache site, and the start-up in 2008. The launching of this project arises from a double finding: 1) the development of nuclear power plants aimed at satisfying the energy needs of the next century cannot be envisaged without the disposal of experimental reactors which are unrivalled for the validation of new concepts of nuclear fuels, materials, and components as well as for their qualification under irradiation. 2) the present park of experimental reactors is 30 to 40 years old and it is advisable to examine henceforth the necessity and the nature of a new reactor to take over and replace, at the beginning of next century, the reactors shut-down in the mean time or at the very end of their lives. Within this framework, the CEA has undertaken, in the last years, a reflection on the mid and long term irradiations needs, to determine the main features and performances of this new reactor. The concept of the reactor will have to fulfil the thermal neutron irradiation requirements as well as the fast neutron experimental needs, with a great potential versatility for any new irradiation programs. The selected reactor project, among several different concepts, is finally a light water open pool concept, with 100 MW thermal power. It could reach neutronic fluxes twice those of present French reactors, and allows many irradiations in the core and around the core, under high neutron fluxes. The reactor will satisfy the highest level of safety in full accordance with international safety recommendations and French safety approach for this kind of nuclear facility, thus giving an added safety margin keeping in mind the versatility of research reactors. The feasibility studies have been focused on the main items, and have permit to determine: the core and fuel designs, with added pressurisation; the different core surrounding structures in connection with the core studies; overall layout of the reactor/auxiliary pools and reactor building. (author)

A series of neutron measurements, for which the principal experimental methods perfected at the Marcoule reactors were used, was carried out at the start-up of EDF1. The measurements were designed mainly to determine the efficiency of the control rods at different depths of insertion. From them a rod-withdrawal configuration was derived which allowed full-power operation without infringing certain limitations on cladding and gas temperatures. At the same time flux measurements were made for different shim-rod positions and different absorber loadings in certain channels. These measurements based on preliminary two-dimensional calculations, were obtained by activation of point detectors,using the standard technique of air poisoning. At certain temperature plateaus (up to 140{sup o}C), measurements of temperature coefficients and control-rod efficiency were made. Spectrum index measurements were carried out at the same time by activation of appropriate detectors (U, Pu, Lu, Mn, In, Au). The oscillation technique was used to measure the efficiency of certain shim rods. Finally, fast-neutron measurements were made in connection with studies of shielding and graphite damage. (author) [French] Une serie de mesures neutroniques utilisant les principales methodes experimentales mises au point sur les reacteurs de Marcoule a ete effectuee au cours du demarrage d'EDF1. Les mesures portent essentiellement sur l 'efficacite des barres de controle a differents enfoncements. On en deduit une configuration de montee des barres permettant d'obtenir la pleine puissance en respectant certaines limitations sur les temperatures de gaines et de gaz. Parallelement des mesures de flux ont ete faites pour differentes positions des barres de compensation et pour divers chargements d'absorbants dans certains canaux, suivant des calculs previsionnels a deux dimensions. Ces mesures sont obtenues par activation de detecteurs ponctuels, au moyen de la technique classique par empoisonnement a l

In order to determine experimentally the amount of radiolysis in the swimming-pool reactor Triton, direct measurements have been made of the quantity of radiolysis gas and hydrogen peroxide in the water, at the entry and exit of the core. The concentration distribution of these gases in the reactor was also determined. An explanation is given as to why no gases evolution is seen in the swimming-pool reactors of the C.E.A. The overall amount of radiolysis is zero, and a simple interpretation of this result is possible. The real amount of radiolysis occurring in the reactor core can be calculated. This is in satisfactory agreement with certain measurement mad elsewhere. (authors) [French] Pour determiner experimentalement le taux de radiolyse dans la pile piscine Triton, des mesures directes de la quantite de gaz de radiolyse et d'eau oxygenee dans l'eau a l'entree et a la sortie du coeur ont ete faites. La repartition de la concentration de ces gaz dans la piscine a egalement ete determinee. On explique pourquoi aucun degagement gazeux n'est observe dans les piles piscines du CE.A. Le taux de radiolyse global est nul, et une interpretation simple de ce resultat est possible. Un taux de radiolyse reel dans le coeur du reacteur peut etre calcule. Celui-ci est en accord satisfaisant avec certaines determinations faites ailleurs. (auteurs)

Since the problems concerning neutron and gamma flux measurements which arise during irradiation experiments in the reactors in the Grenoble and Saclay Centres are of the same type, and since the solutions found are very often adopted in common, we have attempted to describe the methods we use at the present time. A brief description is given of the production of the detectors, the electronic apparatus; the formulae usually used for the interpretation of the measurements are given. A series of technical data cards give the most commonly used detector characteristics. These cards give the physical characteristics of the detectors, their nuclear constants, if any, the most suitable counting methods and the field of application. (authors) [French] Les problemes de mesures de flux de neutrons et de flux gamma qui se posent pour les experiences irradiees dans les reacteurs des Centres de Grenoble et de Saclay etant du meme type et les solutions trouvees, tres souvent adoptees en commun, nous avons cherche a decrire les methodes que nous pratiquons actuellement. On decrit tres brievement la fabrication des detecteurs, l'appareillage electronique; on rappelle les formules usuelles qui servent dans l'interpretation des mesures. Une serie de fiches techniques rassemble les caracteristiques des detecteurs les plus couramment utilises. Ces fiches indiquent les caracteristiques physiques des detecteurs, leurs constantes nucleaires s'il y a lieu, les methodes de comptage les mieux adaptees et le domaine d'utilisation. (auteurs)

The reaction probabilities in an infinite and homogeneous medium of BeO or Be have been calculated from neutron cross-section curves, for a neutron produced with an energy distribution similar to a fission spectrum; the calculation shows that, after several elastic collisions, the neutron has yet an appreciable probability to undergo a reaction, in spite of the energy degradation in the spectrum due to each collision. This degradation has been calculated, taking into account of anisotropy of the collisions. The gain of the reactivity in a reactor has been obtained after correcting these probabilities for the attenuation of the flux of fission neutrons due to the inelastic scattering in the uranium. Finally, the calculation shows that in a power reactor, this gain of reactivity is in practice destroyed in a few years by the accumulation of poisonous nuclei such as Li{sup 6} and He{sup 3} following (n, {alpha}) reaction. (author) [French] Les probabilites de reaction en milieu infini et homogene de glucine (BeO) ou de beryllium ont ete calculees a partir des courbes de section efficace pour un neutron naissant suivant un spectre de fission; le calcul montre qu'apres plusieurs diffusions elastiques le neutron a encore une probabilite appreciable de subir une reaction, malgre la degradation du spectre a chaque diffusion; cette degradation a ete calculee en tenant compte de l'anisotropie du choc. Le gain de reactivite dans un reacteur a ensuite ete obtenu en corrigeant les probabilites en milieu homogene de l'effet l'attenuation du flux des neutrons de fission par les chocs inelastiques dans les barres d'uranium. Enfin, le calcul montre que, dans un reacteur de puissance, ce gain de reactivite est pratiquement detruit en peu d'annees par l'accumulation de noyaux poisons Li{sup 6} et He{sup 3} consecutive a la reaction (n, {alpha}). (auteur)

In a first part, this work deals with the theoretical aspects of the methanization of the industrial effluents; the associated reactional processes are detailed. The second part presents the technological criteria for choosing the methanization process in terms of the characteristics of the effluent to be treated. Some of the methanization processes are presented with their respective advantages and disadvantages. At last, is described the implementation of an industrial methanization unit. The size and the main choices are detailed: the anaerobic reactor, the control, the valorization aspects of the biogas produced. Some examples of industrial developments illustrate the different used options. (O.M.)

This thesis deals with automatic recommendation systems. Automatic recommendation systems are systems that allow, through data mining techniques, to recommend automatically to users, based on their past consumption, items that may interest them. These systems allow for example to increase sales on e-commerce websites: the Amazon site has a marketing strategy based mainly on the recommendation. Amazon has popularized the use of automatic recommendation based on the recommendation function that...

The cans of fuel cartridges used in reactors of the gas-graphite type have either longitudinal fins of variable thickness, short herring-bone fins, or else a mixture of the two. An important test of the strength of these cartridges is their behaviour during thermal cycling carried out in cells reproducing in-pile conditions. It has been observed during with rapid cooling that there occurs a shortening at the base of the fins which can be accompanied in particular by a compression effect at the fin type, which has a tendency to curl, and by a tractive force acting on the body of the can at the ends of the longitudinal fins; this last phenomenon can result in a fracturing of the welds at the extremities or of the ends of the cartridge. This report presents first of all the way in which the stress diagram can be drawn for a can touching the fuel, and then the effect of the ratchet along a fin fixed to a bar with or without grooves. Finally the importance is shown of the test cycling variables (temperature, heating and cooling rates). (authors) [French] Les gaines des cartouches combustibles des reacteurs de la filiere graphite-gaz comportent soit des ailettes longitudinales plus ou moins epaisses, soit de courtes ailettes a chevrons, soit un ensemble des deux. Un test important de la tenue des cartouches, est la tenue au cyclage thermique en cellule pour reproduire le comportement en pile. On a observe au cours des cyclages a refroidissement rapide, un raccourcissement a la base des ailettes qui peut s'accompagner notamment d'une mise en compression du sommet de l'ailette qui a tendance a friser, et d'une traction exercee sur le corps des gaines au bout des ailettes longitudinales; ce dernier phenomene peut se traduire par des ruptures de soudures d'extremites ou des parties terminales de la cartouche. Ce rapport presente d'abord la maniere dont peut etre trace le diagramme des contraintes dans une gaine liee au combustible, puis l'effet du rochet le long d

Fuel-cycle costs and doubling time of fast-breeder reactors are strongly affected by the fuel-burn-up obtainable. Use of oxide or carbide fuel offers the possibility of reaching a burn-up of 100 000 MWd/t. In fuel-clad elements, a limiting factor is the fission-gas-pressure build-up. At the high burn-up considered, an appreciable fraction of the fission gases gets into the pores and thus contributes to the pressure on the can. Starting from the known fission-product yields and decay chains, gas production and pressure build-up have been calculated. Three physical models have been employed in calculating the pressure acting upon the can : the gas is contained either in interconnected pores, in separate pores, or in a central hole. The pressure-dependence upon free volume (fuel density) and temperature will be discussed. Cans made of high-strength materials as Ineonel-X and molybdenum could stand the fission-gas pressure at operating temperatures. Unfortunately, these materials have higher absorption cross-sections than stainless steel. Results of a multi-group calculation are given, showing the effect of using these can materials and of decreasing the fuel density on critical mass and breeding ratio in small and medium-size breeders. (author) [French] Le cout du cycle de combustible et la periode de doublement des reacteurs surgenerateurs a neutrons rapides dependent etroitement du taux de combustion. En utilisant pour combustible un oxyde ou un carbure, on peut atteindre un taux de combustion de 100 000 MW j/t. Avec des combustibles gaines, l'accumulation de la pression des gaz de fission est un facteur limitatif. Pour le fort taux de combustion envisage, une fraction non negligeable des gaz de fission penetre dans les interstices et contribue ainsi a la pression sur la gaine. A partir des rendements en produits de fission et des chaines de desintegration connus, l'auteur a calcule la production de gaz et l'accumulation de pression. Pour calculer la pression

The theory of coupled systems was extensively developed by Avery and co-workers at the Argonne National Laboratory. One of the main points of interest in a coupled system is the larger effective lifetime of neutrons. The effect of the thermal component acts as a sort of neutron-delayer. As in the theory of delayed neutrons the delaying effect disappears if the reactivity worth is high enough to make the fast component critical by itself. In the study a coupled reactor is considered where the fast component suffers a sudden reactivity step {alpha}{sub 0}. Because of the increasing power-level the temperature rises and two temperature coefficients start to work: the temperature coefficient of the fast component and the temperature coefficient of the thermal component. The problem is considered with one group of delayed neutrons (in the ordinary meaning). A formalism is given to express the effective lifetime and temperature coefficient during the different stages of the excursion. Excursions for different {alpha}{sub 0} are given so that the limit of fast-reactor kinetics is reached. (author) [French] La theorie des systemes a couplage a ete mise au point par Avery et ses collaborateurs au Laboratoire national d'Argonne. L'une des caracteristique les plus interessantes d'un systeme a couplage est que le temps de vie effectif des neutrons est plus long. L'effet de la partie thermique contribue en quelque sorte a retarder les neutrons. Comme dans la theorie des neutrons retardes, l'effet de retardement disparait lorsque la reactivite a une valeur suffisamment elevee pour rendre la partie rapide critique par elle-meme. L'auteur du memoire considere un reacteur a couplage dont la partie rapide subit un saut instantane de reactivite, {alpha}{sub 0}. La temperature s'eleve a cause de l'augmentation de puissance et deux coefficients de temperature commencent a s'appliquer: le coefficient de temperature de la partie rapide et le coefficient de temperature de la partie

The high thermal fluxes, which are usual in the latest materials testing reactors, impose suitable paths for uniform heat transfer and a reliable heat removal avoiding bulk-vapour formation. Furthermore, to control the over-all swelling and reactor fuel behaviour, water-gap measurements will also be performed in post-irradiation experiments on spent fuel elements. For that purpose, a probe for measuring the 3-mm water-gap of the BR-2 fuel element over a 1-m length, based on the principle of ultrasonics, has been developed. In the case of post-irradiation experiments, the measuring probe should operate in a fuel element by being immersed in a water pool at a depth of at least 6 m. The probe can withstand prolonged immersion in water and is not affected by normal gamma-irradiation doses. Although operating on the usual pulse-reflection method, the system allows emitted and reflected pulses to be separated by a 10-MHz ferro-electric crystal with high inherent energy dissipation. Oscilloscope read-out can be used, whereby the time is displayed on the horizontal axis, the scanning speed being adjusted to bear a direct relation to the velocity of wave propagation, i.e. the gap distance. This type of read-out Is satisfactory where the number of measurements is restricted, but chart recorder read-out is obviously desirable. In this case, emitted and reflected pulses are shaped and fed to a time-voltage converter using transistor logic techniques. The instrument allows continuous adjustment of output zero for any arbitrary gap distance between 2 and 4 mm thereby permitting zero-centre recording. Furthermore, any desired 100-{mu}m gap distance variation can give a stable 1-V output voltage to a recorder. An accuracy of 5-{mu}m gap-distance variation is easily obtained. Several fuel elements have been measured. The results and reproducibility were very satisfactory. (author) [French] Etant donne que dans les plus recents reacteurs d'essai de materiaux les flux thermiques sont

Material problems associated with the control- and safety-rod systems for graphite moderated, tube-type reactors can be divided into two categories: control materials and operating-channel liner materials. The control materials, such as boron or gadolinium, can be integral with the rod sheath, as in the boron stainless steel used for safety rods. Another approach is the enclosure of a boron-containing sintered compact, such as B{sub 4}C-graphite or B{sub 4}C-aluminium, in a metallic sheath. Rods of the latter type are adaptable for control purposes because of the increased percentages of boron that can be included. Test and fabrication experience indicate that a wide range of satisfactory rod designs is possible with any of these materials. The rod operating channels in the reactor often require liners to protect the surrounding graphite moderator from rod-insertion impact loads and wear and to help maintain channel alignment. Abrasion- and impact resistant, high-strength, low cross-section materials that will operate uncooled are required for these liners. Pyrolytic graphite, pyrolytic graphite composites, aluminium oxide and silicon carbide have been tested for such applications. Physical and irradiation damage data indicate that some of these materials are suitable for lining rod-operating channels. (author) [French] Les problemes de materiaux lies aux dispositifs de controle a barres de reglage et de securite pour les reacteurs tubulaires ralentis au graphite sont doubles et concernent les materiaux absorbants d'une part et les materiaux de garnissage des canaux d'autre part. Les materiaux absorbants tels que le bore ou le gadolinium peuvent former un tout avec le materiau de gainage comme dans le cas ou les barres de securite sont en acier inoxydable au bore. Une autre technique consiste a enfermer un melange presse et fritte contenant du bore, tel que B4C-graphite ou B4C-aluminium, dans une gaine metallique. Les barres de ce dernier type peuvent etre adaptees

electro-magnetic method for technical as well as economic reasons. The optimum area of application of these two methods is explained as well as the large area of overlap where results produced by well- designed and properly operated equipment of both types are essentially equivalent. Spurious defect indications contribute directly to increased component costs, so an evaluation of these effects for both the ultrasonic and the electromagnetic test methods is included for several commonly encountered sources of spurious defect signals. The experience in the application of these methods at Argonne National Laboratory on relatively large quantities of tubing from various sources are recounted from the standpoint of the lowest possible inspection cost per unit length of tubing. This section also summarizes experience gained at Argonne with the newer pulsed electromagnetic test methods. The critical but generally unappreciated role of tube diameter and wall thickness on tube inspection cost is discussed. Since the question of economical inspection is closely related to allowable defect levels, defect levels and standards in use at Argonne are covered. Finally, the practical and theoretical barriers to reduced component inspection costs are enumerated and a projection of what possible reductions in cost might be attainable in the future with the ultrasonic and electromagnetic test methods is attempted. (author) [French] Le reacteur ideal aurait entre autres caracteristiques celle de ne pas exiger de controles non destructifs. Cet ideal, comme tant d'autres, ne sera probablement jamais atteint. Dans l'etude de tout reacteur pour lequel le prix de revient constitue un facteur important, il faudrait envisager la question de savoir si les pieces de ce reacteur pourront etre essayees de facon economique en meme temps que l'on examine les possibilites de fabrication. Cette partie du memoire contient quelques considerations a ce propos ainsi qu'un expose de l'importance des essais non

The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost.

The Jules Horowitz Reactor (RJH) is a new research reactor project dedicated to materials and nuclear fuel testing, the location of which is foreseen at the CEA-CADARACHE site, and the start-up in 2006. The launching of this project originated from a double finding: The development of nuclear power plants aimed at satisfying the energy needs of the next century, cannot be envisaged without experimental reactors which are unrivaled for the validation of new concepts of nuclear fuels, materials, and components as well as for their qualification under irradiation. The existing experimental reactors are 30 to 40 years old and it is advisable to examine henceforth the necessity for and the nature of a new reactor to take over and replace, at the beginning of next century, the reactors shut-down in the mean time or at the very end of their lives. Within this framework, the CEA has undertaken, in the last years, a study on the mid and long term irradiation needs, to determine the main features and performances of this new reactor. The concept of the reactor will have to fulfill the thermal neutron irradiation requirements as well as the fast neutron experimental needs, with a great potential versatility for any new irradiation programs. The reactor project selected among several different concepts, is finally a light water pool concept, with 100 MW thermal power. It could reach neutronic fluxes twice those of present French reactors, and allows for many irradiations in and around the core, under high neutron fluxes. The reactor will satisfy the highest level of safety in full accordance with international safety recommendations and the French safety approach for this kind of nuclear facility, thus giving an added safety margin keeping in mind the versatility of research reactors. The feasibility studies have been focused on the following most important items: neutronic and thermalhydraulic studies on alternative core designs, with or without added pressurization, assessment of different core surrounding structures in connection with the core studies, overall layout of the reactor/auxiliary pools and reactor building. (author)

On April 10. 2007, the french government authorized EDF to create on the site of Flamanville ( La Manche) a nuclear base installation containing a pressurized water EPR type reactor. This nuclear reactor, conceived by AREVA NP and EDF, is the first copy of a generation susceptible to replace later, at least partly, the French nuclear reactors at present in operation.Within the framework of its mission of technical support of the Authority of Nuclear Safety ( A.S.N.), the I.R.S.N. widely contributed successively: to define the general objectives of safety assigned to this new generation of pressurized water nuclear reactors; to analyze the options of safety proposed by EDF for the EPR project; To deepen, upstream to the authorization of creation, the evaluation of the step of safety and the measures of conception retained by EDF that have to allow to respect the objectives of safety which were notified to it. (N.C.)

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

The potential usefulness of dispersions of PuO{sub 2}, UO{sub 2} and ThO{sub 2} in matrices of BeO, Al{sub 2}O{sub 3}, MgO and SiO{sub 2} is reviewed in terms of fuel integrity and fabrication. Dimensional stability and fission-product retentivity are the two features most important to fuel integrity. Compatibility of the constituents of the fuels with one another and with the coolant will influence dimensional stability, but oxide fuels are well favoured in these respects. Dimensional changes under irradiation will contain contributions from neutron and fission fragment damage to the matrix, from radiation damage to the fissile-fertile phase and from agglomerated fission-product gases. Thermal stresses are also capable of effecting changes in shape. However, information on mechanisms for stress relaxation is too limited to enable any reasonable theoretical assessment of behaviour to be made. Both light irradiation and high burn-up studies of fission-product release from the fissile-fertile oxides have concerned themselves mainly with the gaseous products, chiefly xenon. Data on the release of other fission products is very limited as is also information on the movement of fission products in general through the potential matrix materials. Studies of the permeability of sintered pure oxides indicate that densities of at least 95% theoretical density (maybe even 98%) will be needed to eliminate open porosity in such matrices. A variety of techniques are available for the preparation of fissile-fertile particles, for their coating and for their incorporation into high-density matrices. Work on laboratory-scale fabrication processes is well advanced. (author) [French] L'auteur examine la possibilite d'utiliser des combustibles disperses - PuO{sub 2}, UO{sub 2} et ThO{sub 2} et matrices de BeO, Al{sub 2}O{sub 3}, MgO et SiO{sub 2} - dans des reacteurs a haute temperature, au point de vue de l'integrite du combustible et de sa transformation. La stabilite dimensionnelle

The platinoids deposits, currently in exploitation, are often associated with dunitic and gabbroitic ultra-basic rocks containing copper, nickel and iron sulfides. The proportion of platinoids in these deposits is variable and requires generally an adaptation of the industrial processes for their recovery. Nevertheless, we can consider that the classical industrial way consists to treat the composite concentrates of copper-nickel-platinoids sulfides by melting/conversion operations following by hydrometallurgical operations of recovery of the basic metals Cu, Ni. These last ones concentrate the platinoids in leaching residues. In these residues, the platinoids are extracted and separated by hydrometallurgical ways including different techniques: dissolution, selective precipitation, distillation, solvent extraction, resins...The great variety of uses of platinoids makes their recycling difficult. The wastes are often too complex for being economically recycling. The environmental protection depends of the type of industry (metallurgical, chemical, mining). Some products present risks of fire and explosion. The toxicity often appears by superficial allergies and by serious troubles in case of ingestion of soluble salts. (O.M.)

In this dissertation the concept and implementation of a software system supporting the analysis and the design of industrial energy systems is presented. As a basis, a software framework was designed supplying a domain specific object model allowing the description of energy systems as well as the energy auditing projects performed with the software. Moreover, a set of graphical and textual editors needed to model the examined systems is part of the framework. On the other hand, the professional methods for analysis, assessment and optimization of energy systems are implemented in modules integrated into the system via a plug-in interface. The object model whose definition was based on a meta model approach allows the description of network like structures typical to energy systems. In order to keep track of the different work steps performed during an analysis project, these steps are reflected in the object model as ''method applications'' using a tree as the basic structure of a project. In order to allow the compatibility of information a set of conventions for the evaluation of energy flows and system balances was introduced. Moreover, all data elements used in modules or model components are derived from a central database guaranteeing a consistent usage of terms, descriptions, validity ranges and data types. The single professional modules like simulators or optimization methods access the object model via an appropriate software interface. Moreover, they make use of the framework's user interface engine by delivering a generic description of dialog screens and result reports. As all modules share the same set of objects modelling the components of the energy system surveyed, the flow of information from module to module can be designed virtually seamless. Compared to a number of stand-alone solutions, this integrated design approach has the advantage that by combining a set of specialized methods an overall solution for complex engineering tasks can be created. The modules implemented up to now form a solution for the computational support of industrial energy audits. Nevertheless, the framework could be used as a basis for other fields of engineering. As an example for a complex module, a simulator for the calculation of industrial energy supply systems is presented. This module allows modelling of supply systems with low effort in order to calculate the annual costs, system efficiency and emissions. Besides technical components for the conversion, storage and transport of energy, other decisive elements like energy tariffs can be modelled. As input for the simulation, time series of the different target energy demands are needed. As detailed design data of the components is usually not available, the model parameters are typically restricted to the data found in technical data sheets. Moreover, typical sample times of energy demand time series will be 15 minutes or higher, so that dynamic effects below this time interval are neglected. For the purpose of analysis it turns out to be advantageous to assess a supply system without the influences of the concrete system control strategy, which means to run the simulation under the regime of an optimal control strategy. Even in cases where the control strategy is to be taken into account, this approach allows a simpler modelling of the system control as aspects with little impact on the system efficiency can be left to the optimizer instead of formulating appropriate rules. In order to allow an operation optimization of machines whose efficiency depends on temperatures in addition to the part load state (e.g. chillers), an optimization method which allows for quadratic constraints was selected. In order to achieve a method most robust towards the large variety of systems structures to be handled, a combination of evolutionary algorithms and mixed integer linear programming was chosen. In order to create supply system models, component models can be chosen from a template library and combined by connecting their external interfaces with bonds representing an energy or medium flow of a certain type. Each model component may contain subsystems leading to a hierarchical model structure. To define the behaviour of the components, a model description language adapted to the mathematical model was developed. The application of the system is demonstrated in an example of use. (orig.)

The title subject has been evaluated by INDIS (Dutch Bureau for Industrial and Professional Investigations) in order to analyse the demand of the target group and to collect data for the product/market strategical part of the long-range planning for 1989-1993 and 1994-1998. INDIS collected data from 20 interviews with policy-making managers from industrial companies, and 9 interviews with consultancy bureaus, engineering bureaus, contractors, business organizations and the government. These data were supplemented by a literature search. The effectiveness of the NOVEM (Netherlands Agency for Energy and the Environment) has been investigated with regard to the possibilities of developed technologies. This investigation was done by Braun Technoconsult. Conclusions and recommendations are given with regard to energy and environmental projects, the roles of the government, the umbrella organizations and the NOVEM and finally the effectivity of the applicated tools (information, counseling, demonstrations, financing, etc.).

The toxicity and ecotoxicity of airborne particulate matter is determined by its physical features, but also by its chemical composition. The standardised exposure of accumulative bioindicator plants is suggested as an efficient and reliable tool to assess and monitor effects of particulate matter...

The French government has finally opted for agreements negotiated with industrialists to reduce their greenhouse gases emissions, opening thus the way of future emission credits markets. The 'TGAP' (General Taxes on the Pollutant Activities) will thus not be implemented. (O.M.)

Countries with large stock of fissile material and producing large quantity of nuclear pure {sup 235}U and {sup 239}Pu are able to allocate part of the stock to non military research. For countries with low stock of fissile material, all the stock is allocated to military research. An economical and technical solution has to be find to dedicate a part of fissile material to non military research and develop the atomic energy industry. It stated the industrial and economical problems and in particular the choice between the use of enriched fuel with high refining cost or depleted fuel with low production cost. It discusses of four possible utilizations of the natural resources: reactors functioning with pure fissile material ({sup 235}U or {sup 239}Pu) or concentrated material ({sup 235}U mixed with small quantities of {sup 238}U after an incomplete isotopic separation), breeder reactors functioning with enriched material mixed with {sup 238}U or Thorium placed in an appropriate spatial distribution to allow neutrons beam to activate {sup 238}U or Thorium with the regeneration of fissile material in {sup 239}Pu, reactors using natural uranium or low enriched uranium can also produce Plutonium with less efficiency than breeder reactors and the last solution being the use of natural uranium with the only scope of energy production and no production of secondary fissile material. The first class using pure fissile material has a low energy efficiency and is used only by large fissile material stock countries to accumulate energy in small size fuel for nuclear engines researches for submarines and warships. The advantage of the second class of reactors, breeder reactors, is that they produce energy and plutonium. Two type of breeder reactor are considered: breeder reactor using pure fissile material and {sup 238}U or breeder reactor using the promising mixture of pure fissile material and Thorium. Different projects are in phase of development in United States, England and Scotland. The third class of reactor using natural uranium as fuel are presented as a possibility for double-function reactor with the production of plutonium and energy, but the neutron balance is lower than with breeder reactor. One solution is to increase the temperature of functioning but it induces to change the structure materials and moderators. Different solutions are discussed about the utilization of graphite or heavy water as moderators. The last class of reactors using natural uranium and producing only energy is considered by countries with no uranium stock, the energy efficiency and balance, as well as the costs, are then of more importance. Finally, it presented conclusions about the different economic strategies about the industrial development of atomic energy in countries with and without fissile material resources. (M.P.)

This report deals with theoretical analysis of various types of integration of heat pumps in the industry, as well as a demonstration plant that serves the project's practical execution. The report describes the system integration between heat pumps and existing industrial cooling systems. Ammonia plants in industry are estimated to have an allocation of 85%, which is why only an analysis of this type of installation as surplus heat supplier is included in this report. In contrast, heat pumps with both CO{sub 2} and Isobutane as the refrigerant are analysed, since these are the interesting coolants for generating high temperature heat. It can be seen through the project that the combination of heat pump with existing cooling installations may produce favorable situations where the efficiency of the heat pump is extremely high while at the same time electricity and water consumption for the cooling system is reduced. The analysis reflects that CO{sub 2} is preferred over Isobutane in the cases where a high level of temperature boost is desired, whereas Isobutane is preferable at low level of temperature boost. In the demonstration project, the report shows that the heat pump alone has a COP of 4.1, while the achieved COP is 5.5 when by considering the system as a whole. In addition to increased performance the solution profits by having a reduction in CO{sub 2} emissions of 81 tons/year and a saving of 470,000 DKK/year. (LN)

The industrial wastes are made of different kind of wastes: the inert wastes, the banal wastes (municipal wastes), the special wastes containing noxious elements with respect to human health and environment, and the radioactive wastes. Each industry generates its own effluents (sludges from water treatment plants and leachates from rubbish dumps). The main water pollutions are due to the fermentescible organic matters, nitrates and heavy metals from the industrial waste waters. The aim of the public water agencies is to better protect the environment and to give help to the industrialists in the management of their wastes: reduction at the source, selective collection, valorization, transportation and processing. Non-valorizable wastes must be processed: physico-chemical and biological processing (bio-filtering, coagulation-flocculation, membranes and industrial gases), incineration (organic wastes), disposal in class 1 technical burial centres after stabilization (ultimate wastes). Since July 2002, only the ultimate wastes will be disposed off and all class 2 and 3 dumps must have been rehabilitated. This work is divided into 2 parts: part 1 gives a presentation of the different types of industrial wastes and of their management (origin of wastes, effluents, heavy metals, environmental impact, legal aspects, wastes management, valorization). The second part describes the different processes for the treatment of industrial wastes (conventional processes, physico-chemical and biological processes, incineration, tipping, processing of radioactive wastes). (J.S.)

This program is divided in three parts: the context and the regulations, the preparation and the implementation, the tools of prevention and the initiatives and the perspectives.In the first part devoted to the context and regulation are: the context by the Authority of nuclear safety (A.S.N.), the regulation referential, the transport of gamma-graphs; in the second part are the distribution of liabilities, materials and associated requirements, the feedback of incidents and exploitation of it, training and base requirements, works of S.F.R.P./C.O.F.R.E.N.D. and the A.S.N. position; the third part includes help to evaluation of risks at working places of industry radiologists, dosimetry study of a working place, guide to evaluate oneself; the fourth part devoted to the initiatives and the perspectives are: regional experiences charters of good practices in industry radiography, integration of works and deployment by the members of the C.O.F.R.E.N.D., perspectives in matter of prevention of occupational risks in the area of industry radiography. (N.C.)

This paper addresses the subject 'stress' and its effects on the personality. Specific types of behavior, which the human develops in a fearful situation are demonstrated by means of the psychoanalytical personality model. In the conclusion possible methods of alleviating fear, shock and panic reactions are described. 7 figs., 2 tabs., 9 refs

Characterization of two industrial effluents in Togo : environment impact study. Environment pollution due to two industrial effluents has been investigated. Results how that effluent derive from the factory of the treatment of Kpémé posphate ore was loaded with settling suspended matter (> 90 % of total suspended solids).

In four articles in this special issue of the magazine attention is paid to new techniques by which emissions to the air from the industry can be controlled and/or prevented. In the first article an overview is given of sources of air pollution, caused by dust. In the second article intermediate results of the KWS 2000 program (aimed at 50% reduction of the emission of volatile organic matter for the year 2000) are outlined. In the third article a cooperative biological (flue) gas desulfurization pilot plant project is discussed. In the fourth and last article the most important possible techniques to reduce the emission of volatile organic matter are highlighted

More than 350 deposits from mining and processing in Baden-Wuerttemberg and Lower Saxony were visited, measured radiometrically, assessed by a unified scheme and partly sampled. The highest local dose rates, up to 100 mSv/a, were found in the waste from uranium exploration, as expected. The values of these planned wastes were followed by waste from silver-cobalt mining dumped above ground, with dose rates up to 20 mSv/a. Fly ash and slag deposits had a dose rate of about 2 mSv/a, and are comparable with many porphyry and granite rocks. The higher values of slag deposits from iron ore smelting were clearly due to high thorium content. The lowest doses were from stone waste from the mining of hydrothermal lead-zinc and iron ores. The analyses of uranium, radium 226 and lead 210 confirmed the radiometric measurements. The analysis of waste seepage water and tunnel outlets gave only a low number of values which exceeded the drinking water concentration according to the Radiation Protection Order. (orig./RB) [de

This work deals with electrostatic precipitators or ESP used for gas-solid particles separation. By means of a dust-controlled testing loop created and realised at the GRETh`s plate-form (Research Group on Heat Exchangers) and a numerical model developed during this work from TRIO software, the study of the performances of different ESP geometries has been carried out. Many electrical, hydraulic and particular parameters governing solid particles collection under ionised electric field have been identified, measured and modelled. The numerical model, ratified with experimental data obtained during this study and from literature, allows to describe local and global phenomena occurring in any geometries. Furthermore, parametric studies have been carried out in order to propose some optimised geometries. allowing to increase collection efficiencies. At least, on-site measurements with CETIAT (Centre Technique des Industries Aerauliques et Thermiques) allow to identify dust particles likely to be thrown out to the atmosphere, and troubles peculiar to large scales industrial plants. The numerical model has also been tested on these data. At the end of this study, an efficient dust-controlled experimental tool, PACIFIC loop, and a numerical simulation allowing ESP sizing are available. (author)

The electrical power quality is a strong criterion for many manufacturers. The improvement of the electrical network and devices has allowed to reach a good power quality level all over the French territory which satisfies the majority of most customer's expectations. However, in some cases, only local desensitization can ne efficient enough to meet the customer's requirement. Emergent equipment based on power electronic devices are evaluated at the Research and Development Division. The aim is to propose the best service offer to our customers on both technical and economical aspects. (authors)

The first two natural uranium-graphite-C0{sub 2} reactors at the Chinon station have metal vessels of thick manganese-molybdenum steel plate. The studies carried out on these vessels raised certain problems, particularly in connection with the design and dimensions of the port reinforcements. The reinforcements for the control-rod channels and fuel ports were studied on mock-ups and the results obtained were checked on the completed reactors during hydraulic tests. The type of construction initially used for the EDF1 vessel was relatively simple. The plates to be welded were locally preheated, and the vessel was not supposed to undergo more than one stress-relief heat treatment after completion of all the welding. Serious cracks developed, however, and it became necessary to alter the whole method of construction. In particular, the welding was now done after overall preheating and the vessel was subjected to multiple stress-relief treatments. This made it possible to fabricate the vessels for EDF1 and EDF2, but at the same time imposed certain limitations which considerably complicated work on the site. (author) [French] Les reacteurs a uranium naturel, graphite et gaz carbonique des deux premieres tranches de la Centrale de Chinon comportent des caissons metalliques realises a partir de toles de fortes epaisseurs, en acier au manganese-molybdene. Les etudes de ces paissons ont pose certains problemes, notamment en ce qui concerne les renforts d'ouvertures. Les renforts des passages des barres de controle et des orifices de chargement ont ete etudies sur maquette et les resultats obtenus ont ete controles sur les ouvrages termines lors des epreuves hydrauliques. Le mode de construction initialement utilise pour le caisson de la tranche EDF1 etait relativement simple; les toles a souder etaient prechauffees localement et le caisson ne devait subir qu'un seul traitement thermique de detente, apres execution de toutes les soudures. Une fissuration importante en cours

This document proposes an approach of the global cost of an industrial lightning installation and provides information on the benefits and its associated performance of a good industrial lightning, examples of applications, the available products and the regulation texts of the domain. (A.L.B.)

The use of nuclear reactor for energy production induces the installation of automatic control systems which need to be safe enough and can adapt to the industrial scale of energy production. These automatic control systems have to insure the constancy of power level and adjust the power produced to the energy demand. Two functioning modes are considered: nuclear plant connected up to other electric production systems as hydraulic or thermic plants or nuclear plants functioning on an independent network. For nuclear plants connected up with other production plants, xenon poisoning and operating cost lead to keep working at maximum power the nuclear reactors. Thus, the power modulation control system will not be considered and only start-up control, safety control, and control systems will be automated. For nuclear power plants working on an independent network, the power modulation control system is needed to economize fuel. It described the automated control system for reactors functioning with constant power: a power measurement system constituted of an ionization chamber and a direct-current amplifier will control the steadfastness of the power produced. For reactors functioning with variable power, the automated power control system will allow to change the power and maintain it steady with all the necessary safety and will control that working conditions under P{sub max} and R{sub max} (maximum power and maximum reactivity). The effects of temperature and xenon poisoning will also be discussed. Safety systems will be added to stop completely the functioning of the reactor if P{sub max} is reached. (M.P.)

The aim of this study is to develop techniques to qualify the efficiency of liquid distribution in trickle-bed reactors, using cold mockups. The experimental setup consists mainly in a 0.3-m-ID packed-bed column with three different plates used to vary the quality of inlet liquid distribution. Liquid distribution has been qualified using several techniques: global pressure drop measurements, global RTD (Residence-Time Distribution) of the liquid, local heat transfer probes, capacitance tomography, collector at the bottom of the reactor with nine equal zones. The bed pressure drop and the overall external liquid saturation decrease when the maldistribution increases; quantitative information is however difficult to obtain this way. Global RTD of the liquid allows quantifying of the average liquid distribution in the bed. The local thermal sensors give an indication of local liquid velocity and indicate possible local maldistribution of the liquid (scale mm) even when global distribution is good. Concerning the results obtained with the collector, a maldistribution index is defined ranging from 0 (ideal distribution) to 1 (worst possible distribution), and the influence of the different operating parameters (gas and liquid velocities, particle shape) is discussed. (authors)

The SFEN organized, the 10 june 1999 at Paris, a meeting in the domain of the PWR hydraulics and in particular the hydraulic phenomena concerning the vessel and the vapor generators. The papers presented showed the importance of the industrial stakes with their associated phenomena: cores performance and safety with the more homogenous cooling system, the rods and the control rods wear, the temperature control, the fluid-structure interactions. A great part was also devoted to the progresses in the domain of the numerical simulation and the models and algorithms qualification. (A.L.B.)

This work chiefly concerns the measurement of fast neutron fluxes by means of threshold detectors. It is shown first that the cross sections to use for measurements by threshold detectors depend largely on the neutron spectrum, that is the position in which the measurement is performed. The spectrum is determined by calculation for several positions in the piles EL2 and EL3; from this can be deduced the cross-sections to be used for the measurements carried out in these positions. In the last part of the report, possible methods for the experimental determination of the spectrum are indicated. (author) [French] On etudie principalement la mesure des flux de neutrons rapides a l'aide de detecteurs a seuil. On montre d'abord que les sections efficaces a utiliser pour les mesures par detecteurs a seuil, dependent grandement du spectre des neutrons, c'est-a-dire de l'emplacement ou s'effectue la mesure. La determination du spectre est effectuee par le calcul pour plusieurs emplacements des piles EL2 et EL3; on en deduit les sections efficaces a utiliser pour les mesures effectuees a ces emplacements. Dans la derniere partie du rapport, on indique quelles methodes sont possibles pour la determination experimentale du spectre. (auteur)

Resonance self-shielding calculations are an essential component of a deterministic lattice code calculation. Even if their aim is to correct the cross sections deviation, they introduce a non negligible error in evaluated parameters such as the flux. Until now, French studies for light water reactors are based on effective reaction rates obtained using an equivalence in dilution technique. With the increase of computing capacities, this method starts to show its limits in precision and can be replaced by a subgroup method. Originally used for fast neutron reactor calculations, the subgroup method has many advantages such as using an exact slowing down equation. The aim of this thesis is to suggest a validation as precise as possible without burnup, and then with an isotopic depletion study for the subgroup method. In the end, users interested in implementing a subgroup method in their scheme for Pressurized Water Reactors can rely on this thesis to justify their modelization choices. Moreover, other parameters are validated to suggest a new reference scheme for fast execution and precise results. These new techniques are implemented in the French lattice scheme SHEM-MOC, composed of a Method Of Characteristics flux calculation and a SHEM-like 281-energy group mesh. First, the libraries processed by the CEA are compared. Then, this thesis suggests the most suitable energetic discretization for a subgroup method. Finally, other techniques such as the representation of the anisotropy of the scattering sources and the spatial representation of the source in the MOC calculation are studied. A DRAGON5 scheme is also validated as it shows interesting elements: the DRAGON5 subgroup method is run with a 295-eenergy group mesh (compared to 361 groups for APOLLO2). There are two reasons to use this code. The first involves offering a new reference lattice scheme for Pressurized Water Reactors to DRAGON5 users. The second is to study parameters that are not available in APOLLO2 such as self-shielding in a temperature gradient and using a flux calculation based on MOC in the self-shielding part of the simulation. This thesis concludes that: (1) The subgroup method is at least more precise than a technique based on effective reaction rates, only if we use a 361-energy group mesh; (2) MOC with a linear source in a geometrical region gives better results than a MOC with a constant model. A moderator discretization is compulsory; (3) A P3 choc law is satisfactory, ensuring a coherence with 2D full core calculations; (4) SHEM295 is viable with a Subgroup Projection Method for DRAGON5.

Accelerator driven sub-critical reactors are subject of many research programmes since more than ten years, with the aim of testing the feasibility of the concept as well as their efficiency as a transmutation tool. Several key points like the accelerator, the spallation target, or neutronics in a subcritical medium were investigated extensively these last years, allowing for technological choices and the design of a low power European demonstration ADS (a few tens of MWth). Programmes dedicated to subcritical reactor piloting proposed a monitoring procedure to be validated in forthcoming experiments. Accelerator R and D provided the design of a LINAC for an ADS and research work on accelerator reliability is going on. A spallation target was operated at PSI and the design of a windowless target is in progress. All this research work converges to the design of a European demonstration ADS, the ETD/XT-ADS, which could be the Belgian MYRRHA project. (author)

In the framework of the FMDP (Fissile Materials Disposition Program between the US and Russian, a benchmark was tested. The pin cells contain low enriched uranium (LEU) and mixed oxide fuels (MOX). The calculations are done for a wide range of temperatures and solute boron concentrations, in accidental conditions. (A.L.B.)

International audience; Price war on the shelves for national and international brands, pursuit of food retail groups concentration or supermarket procurement centres concentration, cartels in the industrial sector to cope with retailers purchasing power, French Competition Authority’s recent reports, ALUR Law (2014) and Macron Law (2015) including some policy on retail ... news shows that vertical competition between the food industry and the large retailers is exacerbated in a context of we...

The total common industrial waste volume production in France has been evaluated, taking into consideration all the industrial and commercial sectors and the following materials: glass, metals, plastics, rubber, textiles, papers, cardboard, wood, leather, organic matters, building wastes, mixtures. Results are presented for the various regions of France, as a function of enterprise size, waste type and destination; data are also given concerning packaging materials, and waste collection and processing. Comparisons are made with data from other information sources and calculations

Increasing energy efficiency is one of the main challenges in the plastics processing industry. However, the structure and the parameters of the cooling sections of extrusion lines have been optimized only in a few cases with respect to maximum efficiency and minimal operating costs. Thus, an innovative cascaded cooling section enables not only an energy-efficient cooling, but also an improvement of the properties of the finished plastic products.

This document provides information on the systems of possibilities of power variation and their management for an industrial lightning in the ternary sector. These systems bring to the user more comfort and improve the global and environmental cost of the installations. (A.L.B.)

The French industrial production of nanocrystalline precursor ribbon (Imphy Ugine Precision - IUP) and nanocrystalline wound cores (Mecagis) is now mature, promoting then one of the first worldwide provider in this market. Recent progress in ribbon elaboration will provide large increase of industrial efficiency, leading the cost of a nanocrystalline solution to be closed to the cost of a ferrite solution. The precise study and control of magnetoelastic energy allowed the production scattering to be reduced, the alloy to be weakly dependant on external stresses (production, packaging, thermal dilatation), further promoting the performances. Whatever the alloy is very brittle in the nanocrystalline state, some improvements are using or are going around this intrinsic behaviour, and are now developed: powder core for low dissipative filtering, cut core for storage and strong power transformation, wound cores from ribbon nano-crystallized with high stresses during annealing, for the storage and current metering. (authors)

This paper presents an analysis of safety standards and their implementation in certification strategies from different domains such as aeronautics, automation, automotive, nuclear, railway and space. This work, performed in the context of the CG2E ('Club des Grandes Entreprises de l'Embarque'), aims at identifying the main similarities and dissimilarities, for potential cross-domain harmonization. We strive to find the most comprehensive 'trans-sectorial' approach, within a large number of industrial domains. Exhibiting the 'true goals' of their numerous applicable standards, related to the safety of system and software, is a first important step towards harmonization, sharing common approaches, methods and tools whenever possible. (authors)

In an increasingly fragmented gas market, the focus for national gas safety may not rest with one major utility or gas supplier but may be spread across many companies. There will also be many new organisations in a liberalized gas industry with varying views on the needs and benefits of safety related technology development but all agree there is a need to ensure that the good safety record of gas as a domestic fuel is maintained. The number of carbon monoxide (CO) incidents is not decreasing significantly despite an increased awareness of the problem. As a consequence, a two-year joint industry programme addressing issues related to carbon monoxide has been established, co-ordinated by BG Technology and supported by gas organisations, government agencies, manufacturers and suppliers across Europe and the World. The 2-year 2 pound million programme has been constructed as twelve separate projects addressing issues such as the reporting and analysis of domestic incidents, improved service or installation practice, CO alarm reliability and information dissemination. The paper gives results and achievements of the programme, through new techniques, standards, procedures or equipment and demonstrates how the gas industry can work together to meet common safety objectives. (authors)

The various heat transfer equations used for calculations in thermal diagnosis of an industrial building are reviewed: calculation of the heat losses through walls as a function of building materials, calculation of the energy consumption for heating fresh air (as a function of the air pollution rate in the building), calculation of the total heat losses, the heating energy demand and the annual energy consumption. Data concerning building materials characteristics, insulation and heating loads in the various regions of France, are also presented

The report assesses whether industrial exploitation of gas from possible new major discoveries outside Northern Norway could be profitable. Profits are uncertain and depend heavily on gas prices. Industrial exploitation depends on the development of large, capital-intensive plants. This increases the financial risk. In a situation where gas export through pipelines is not an alternative industrial exploitation could be an alternative to export via LNG. However, the pressing demand for natural gas increases the probability of the construction of a gas export pipeline as a realistic alternative. Carbon emissions from a land-based plant will be important, and the costs for trading of emission allowances will reduce profitability. The authorities should allow for industrial exploitation to be assessed on an equal basis with pipeline transport and LNG as alternatives for market solutions for major new discoveries. (EW)

The electro-coagulation is a water treatment technic in electrolysis cell with double anode. In substitution to the coagulant reagent often used in water de-pollution, it realizes also the coloring decomposition, the DCO abatement and sometimes improving the sludges processing. The technic presents meanwhile some limitations as its poor treatment capacity and the necessity of a high effluent conductivity. An example of application shows that this technic is economically competitive. (A.L.B.)

The IR dryer is used in a paper making machine to dry the paper web after it has been coated with a surface layer. In part 1 of the project a mathematical model have been developed, capable of calculating the radiation intensity and other energy flows in the dryer. In part 2 of the project, measurements have been made on the IR radiator mounted in the paper making machine. The calculation model shows the efficiency of the radiator to 39% at full power and 35% at half power. The direct measurements were made at half power and gave an efficiency of 31% for new radiators and 28% for old ones. The conclusion is that the calculation model values corresponds very well compared with direct measurements.

Vérification of PLC (Programmable Logic Controller) programs is important when these programs are to control critical applications for reactive systems. This context has already been study for untimed programs. In this thesis, We are interested to timed properties and programs. More precisely, we give a formal semantics to (partial) Ladder diagrams and TON blocks, with timed automata. We also propose a timed logic in order to abstract transient events, where transient properties is parameteri...

The contribution presents the software package 'TOP-Energy' which supports energy consultants in their analysis and optimisation of industrial energy systems and is a tool for development and assessment of measures for reducing the energy cost and the consumption of energy resources. In particular, it supports data acquisition, evaluation, and presentation of results of routine work; it offers simulations of complel processes and systems as well as tools like integrated project management. TOP-Energy consists of several modules linked by a common framework. The framework is for data management, module integration and control, and offers a user interface in the form of adaptable editors, dialogues and menus. Power supply systems of industrial works can be modelled with all their components. The key module of Top-energy is a simulator for systems designed, with variable temporal load curves and other boundary conditions. (orig.)

The subject of radioactive wastes is developed through their origin, their classification, their scale of size. The different storage centers are given and the new channels of radioactive wastes management are tackled. The particular case of high level and long term radioactive wastes is detailed. (N.C.)

From the point of view of the industry there are two central tasks for energy policy: 1. The German industry has to be supplied with energy at internationally competitive prices. Of the three aims of energy policies (secure, moderately-priced and low-pollutant) the price aim is in Germany in the international comparison today least achieved. 2. Policy always tends to solve problems in the energy sector with regulating interventions. An example from the recent past is the power supply law of 1990 with which the way to political electricity pricing was opened. But the precept for the Internal European Market has to be deregulation as only in this way the larger market can develop its efficiency to the full. Also for the reform of the economic laws concerning the energy sector these two central aims have to be taken into consideration. (orig./UA).

The increase in production of goods over the last decades has led to the need for improving the management of natural resources management and the efficiency of processes. As a consequence, heat integration methods for industry have been developed. These have been successful for the design of new plants: the integration principles are largely employed, and energy intensity has dramatically decreased in many processes. Although progress has also been achieved in integration methods for retrofit, these methods still need further conceptual development. Furthermore, methodological difficulties increase when trying to retrofit heat exchange networks that are closely interrelated to water networks, such as the case of pulp and paper mills. The pulp and paper industry seeks to increase its profitability by reducing production costs and optimizing supply chains. Recent process developments in forestry biorefining give this industry the opportunity for diversification into bio-products, increasing potential profit margins, and at the same time modernizing its energy systems. Identification of energy strategies for a mill in a changing environment, including the possibility of adding a biorefinery process on the industrial site, requires better integration methods for retrofit situations. The objective of this thesis is to develop an energy integration method for the retrofit of industrial systems and the transformation of pulp and paper mills, ant to demonstrate the method in case studies. Energy is conserved and degraded in a process. Heat can be converted into electricity, stored as chemical energy, or rejected to the environment. A systematic analysis of successive degradations of energy between the hot utilities until the environment, through process operations and existing heat exchangers, is essential in order to reduce the heat consumption. In this thesis, the "Bridge Method" for energy integration by heat exchanger network retrofit has been developed. This method is the first that considers the analysis of these degradations. The fundamental mechanism to reduce the heat consumption in an existing network has been made explicit; it is the basis of the developed method. The Bridge Method includes the definition of "a bridge", which is a set of modifications leading to heat reduction in a heat exchanger network. It is proven that, for a given set of streams, only bridges can lead to heat savings. The Bridge Method also includes (1) a global procedure for heat exchanger network retrofit, (2) a procedure to enumerate systematically the bridges, (3) "a network table" to easily evaluate them, and (4) an "energy transfer diagram" showing the effect of the two first principles of thermodynamics of energy conservation and degradation in industrial processes in order to identify energy savings opportunities. The Bridge Method can be used for the analysis of networks including several types of heat transfer, and site-wide analysis. The Bridge Method has been applied in case studies for retrofitting networks composed of indirect-contact heat exchangers, including the network of a kraft pulp mill, and also networks of direct-contact heat exchangers, including the hot water production system of a pulp mill. The method has finally been applied for the evaluation of a biorefinery process, alone or hosted in a kraft pulp mill. Results show that the use of the method significantly reduces the search space and leads to identification of the relevant solutions. The necessity of a bridge to reduce the inputs and outputs of a process is a consequence of the two first thermodynamics principles of energy conservation and increase in entropy. The concept of bridge alone can also be used as a tool for process analysis, and in numerical optimization-based approaches for energy integration.

This catalogue gives information on 68 chemical, mechanical, magnetic, electrical, thermal, etc. techniques for the processing of solid, viscous and liquid, common or special, industrial wastes. The various processes are presented as files, which are easily retrievable through keywords, waste type or industry codes, processing types, distributors. Technologies, performances and applications of each techniques are presented, together with references and company contacts

In a power reactor destined to supply industrially recoverable thermal energy, the most economical source of heat still consists of natural metallic uranium. However, the nuclear fuel material, most often employed in the form of rods of 20 to 40 mm diameter, is subjected to a series of stresses which lead to irreversible distortions usually incompatible with the substructure of the reactor. As a result the fuel material must possess at the outset a certain number of qualities which must be determined. Investigations have therefore been carried out, first on the technological characters peculiar to each of the three allotropic phases of pure uranium metal, and on their interactions on the stabilisation of the material which consists of either cast uranium or uranium pile-treated in the {gamma} phase. (author) [French] Dans un reacteur de puissance destine a fournir de l'energie thermique industriellement recuperable, la source de chaleur la plus economique reste constituee par de l'uranium metallique naturel. Or, le materiau combustible nucleaire, employe le plus souvent sous forme de barreaux de 20 a 40 mm de diametre, se trouve soumis a un ensemble de contraintes qui provoque des deformations irreversibles, le plus souvent incompatibles avec l'infrastructure du reacteur. Par consequent, le materiau combustible doit presenter a l'origine un certain nombre de qualites qu'il est necessaire de determiner. Aussi a-t-on d'abord etudie les caracteres technologiques propres a chacune des trois phases allotropiques de l'uranium-metal pur et leurs interactions sur la stabilisation du materiau constitue soit par de l'uranium coule, soit par de l'uranium traite en pile en phase {gamma}. (auteur)

In a power reactor destined to supply industrially recoverable thermal energy, the most economical source of heat still consists of natural metallic uranium. However, the nuclear fuel material, most often employed in the form of rods of 20 to 40 mm diameter, is subjected to a series of stresses which lead to irreversible distortions usually incompatible with the substructure of the reactor. As a result the fuel material must possess at the outset a certain number of qualities which must be determined. Investigations have therefore been carried out, first on the technological characters peculiar to each of the three allotropic phases of pure uranium metal, and on their interactions on the stabilisation of the material which consists of either cast uranium or uranium pile-treated in the {gamma} phase. (author) [French] Dans un reacteur de puissance destine a fournir de l'energie thermique industriellement recuperable, la source de chaleur la plus economique reste constituee par de l'uranium metallique naturel. Or, le materiau combustible nucleaire, employe le plus souvent sous forme de barreaux de 20 a 40 mm de diametre, se trouve soumis a un ensemble de contraintes qui provoque des deformations irreversibles, le plus souvent incompatibles avec l'infrastructure du reacteur. Par consequent, le materiau combustible doit presenter a l'origine un certain nombre de qualites qu'il est necessaire de determiner. Aussi a-t-on d'abord etudie les caracteres technologiques propres a chacune des trois phases allotropiques de l'uranium-metal pur et leurs interactions sur la stabilisation du materiau constitue soit par de l'uranium coule, soit par de l'uranium traite en pile en phase {gamma}. (auteur)

One of the methods of producing carrier free phosphorus-32 today is by extraction into water from pile-irradiated sulphur. The present work gives information concerning the chemical states of P{sup 32} in aqueous solutions at different steps of the routine production-process. The variation in the chemical state of P{sup 32} compounds in the final product has also been examined as a function of storage time. P{sup 32} bound as orthophosphate was found to be the main component. During the chemical processing, the amount of orthophosphate increased from about 70% at the beginning of the extraction to about 98 % in the final carrier-free P{sup 32} product. The residual amount consisted of a mixture of pyro-, tri-, tetra-, and other long-chain polyphosphates (number of P {>=} 5). No metaphosphates (ring-formed) were found in the solutions during production and storage. The results indicate that the polyphosphorus compounds were formed in the target material during irradiation. Special attention was paid to the adsorption of carrier-free P{sup 32} compounds to glassware under the existing experimental conditions. (author) [French] L'une des methodes employee a l'heure actuelle pour obtenir du phosphore-32 sans entraineur consiste a l'extraire dans l'eau a partir de soufre irradie dans un reacteur. Les auteurs donnent des indications sur l'etat chimique du phosphore-32 dans des solutions aqueuses, a differentes etapes du processus de preparation courant. Us examinent aussi les changements de l'etat chimique des composes du phosphore-32 dans le produit final en fonction de la duree de stockage. On a constate que le {sup 32}P combine sous forme d'orthophosphate etait le principal composant. Au cours du traitement chimique, la teneur en orthophosphate est passee d'environ 70% au debut de l'extraction a environ 98% lors de l'obtention du produit final sans entraineur. Le reste etait constitue d'un melange de pyro-, tri-, tetra- et autres polyphosphates a chaine longue (P

This report describes one study which contributed to the construction of the Monts d'Arree nuclear power station: EL-4. The reactor is cooled by a CO{sub 2} current provided by 3 turbo-blower groups. The priming vapour for the turbines is taken at the exit of the main CO{sub 2} - H{sub 2}O exchangers. The operation of EL 4 is based on a high degree of centralization of the controls which attributes an important role to the general regulation circuits. This general regulation includes in particular an internal blowing loop which controls the CO{sub 2} flow. The study of the control of this CO{sub 2} flow is made up of 3 parts: - analogue representation of the reactors cooling circuit and of the turbo blower unit. - first test campaign using the analogue computer describing the natural behaviour of the system in the absence of control. theoretical determination of the regulation factors; definition of the regulation using an analogue computer and second test campaign for recording the performances of the blowing loop. The 4. part of the report deals with the analogue study: analogue equations - development. (authors) [French] Ce rapport prend place parmi les etudes de realisation de la Centrale des Monts d'Arree EL-4. Le reacteur est refroidi par une circulation de CO{sub 2} assuree par 3 groupes turbosoufflantes. La vapeur d'entrainement des turbines est prelevee a la sortie des echangeurs principaux CO{sub 2} - H{sub 2}O. L'exploitation de EL-4 repose sur une centralisation poussee des moyens de controle-commande qui attribue un role essentiel aux circuits de regulation generale. Cette regulation generale comporte en particulier une boucle interne de soufflage qui realise un asservissement du debit de CO{sub 2}. L'etude de cette regulation du debit CO{sub 2} comprend 3 parties: - representation analogique du circuit de refroidissement du reacteur et de l'ensemble turbine-soufflante. - premiere campagne d'essais sur calculateur analogique decrivant le comportement

France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [French] La France, qui a base son effort principal pour la production d'energie nucleaire sur la filiere des reacteurs a uranium naturel et graphite refroidis par gaz, et qui a un programme a plus

Heat transfer studies of electrical cable insulation in superconducting winding are of major importance for stability studies in superconducting magnets. This work presents an experimental heat transfer study in superconducting cables of Large Hadron Collider dipoles cooled by superfluid helium and submitted to volume heat dissipation due to beam losses. For NbTi magnets cooled by superfluid helium the most severe heat barrier comes from the electrical insulation of the cables. Heat behaviour of a winding is approached through an experimental model in which insulation characteristics can be modified. Different tests on insulation patterns show that heat transfer is influenced by superfluid helium contained in insulation even for small volume of helium (2 % of cable volume). Electrical insulation can be considered as a composite material made of a solid matrix with a helium channels network which cannot be modelled easily. This network is characterised by another experimental apparatus which allows to study transverse and steady-state heat transfer through an elementary insulation pattern. Measurements in Landau regime ({delta}T{approx}10{sup -5} to 10{sup -3} K) and in Gorter-Mellink regime ({delta}T>10{sup -3} K) and using assumptions that helium thermal paths and conduction in the insulation are decoupled allow to determine an equivalent channel area (10{sup -6} m{sup 2}) and an equivalent channel diameter (25 {mu}). (author)

The emergence of uncooled detectors has opened new opportunities for IR detection for both military and commercial applications. Development of such devices involves a lot of trade-offs between the different parameters that define the technological stack. These trade-offs explain the number of different architectures that are under worldwide development. The key factor is to find a high sensitivity and low noise thermometer material compatible with silicon technology in order to achieve high thermal isolation in the smallest area as possible. Ferroelectric thermometer based hybrid technology and electrical resistive thermometer based (microbolometer) technology are under development. LETI and ULIS have chosen from the very beginning to develop first a monolithic microbolometer technology fully compatible with commercially available CMOS technology and secondly amorphous silicon based thermometer. This silicon approach has the greatest potential for reducing infrared detector manufacturing cost. After the development of the technology, the transfer to industrial facilities has been performed in a short period of time and the production is now ramping up with ULIS team in new facilities. LETI and ULIS are now working to facilitate the IRFPA integration into equipment in order to address a very large market. Achievement of this goal needs the development of smart sensors with on-chip advanced functions and the decrease of manufacturing cost of IRFPA by decreasing the pixel pitch and simplifying the vacuum package. We present in this paper the technology developed by CEA/LETI and its improvement for being able to designs 384×288 and 160×120 arrays with a pitch of 35 μm. Thermographic application needs high stability infrared detector with a precise determination of the amount of absorbed infrared flux. Hence, infrared detector with internal temperature stabilized shield has been developed and characterized. These results will be presented. To cite this article: J.-L. Tissot, C. R. Physique 4 (2003).

Heat transfer studies of electrical cable insulation in superconducting winding are of major importance for stability studies in superconducting magnets. This work presents an experimental heat transfer study in superconducting cables of Large Hadron Collider dipoles cooled by superfluid helium and submitted to volume heat dissipation due to beam losses. For NbTi magnets cooled by superfluid helium the most severe heat barrier comes from the electrical insulation of the cables. Heat behaviour of a winding is approached through an experimental model in which insulation characteristics can be modified. Different tests on insulation patterns show that heat transfer is influenced by superfluid helium contained in insulation even for small volume of helium (2 % of cable volume). Electrical insulation can be considered as a composite material made of a solid matrix with a helium channels network which cannot be modelled easily. This network is characterised by another experimental apparatus which allows to study transverse and steady-state heat transfer through an elementary insulation pattern. Measurements in Landau regime ({delta}T{approx}10{sup -5} to 10{sup -3} K) and in Gorter-Mellink regime ({delta}T>10{sup -3} K) and using assumptions that helium thermal paths and conduction in the insulation are decoupled allow to determine an equivalent channel area (10{sup -6} m{sup 2}) and an equivalent channel diameter (25 {mu}). (author)

The direct conversion of thermal energy at high temperature to electrical energy using plasma diodes is studied in the Electronic Physics Service of the French Atomic Energy Commission. Research concerns the adaptation of these diodes to nuclear reactors with a view especially to providing power for operating the instruments of space vehicles. In parallel with the semi-industrial realizations and tests, an important part of the activity of the service is directed towards fundamental research into physical phenomena convected with thermionic emission with a view to improving present performances. (authors) [French] La conversion directe de l'energie thermique a haute temperature en energie electrique par diodes a plasma est etudiee au Service d'Electronique Physique du Commissariat a l'Energie Atomique. On etudie l'adaptation de ces diodes aux reacteurs nucleaires, notamment en vue de fournir l'energie de servitude dans les vehicules spatiaux. Parallelement aux realisations et essais semi-industriels, une part importante de l'activite du Service est orientee vers l'etude fondamentale des phenomenes physiques lies a la conversion thermoionique dans le but d'ameliorer les performances actuelles. (auteurs)

Supercritical water oxidation is an innovative process to treat organic liquid waste which uses supercritical water properties to mix efficiency the oxidant and the organic compounds. The reactor is a stirred double shell reactor. In the step of adaptation to nuclear constraints, the computational fluid dynamic modeling is a good tool to know required temperature field in the reactor for safety analysis. Firstly, the CFD modeling of tubular reactor confirms the hypothesis of an incompressible fluid and the use of k-w turbulence model to represent the hydrodynamic. Moreover, the EDC model is as efficiency as the kinetic to compute the reaction rate in this reactor. Secondly, the study of turbulent flow in the double shell reactor confirms the use of 2D axisymmetric geometry instead of 3D geometry to compute heat transfer. Moreover, this study reports that water-air mixing is not in single phase. The reactive turbulent flow is well represented by EDC model after adaptation of initial conditions. The reaction rate in supercritical water oxidation reactor is mainly controlled by the mixing. (author)

This paper is a series of transparencies presenting a comparative study of the thermal performances of different types of refrigerating machineries: di-thermal with vapor compression, tri-thermal with moto-compressor, with ejector, with free piston, adsorption-type, resorption-type, absorption-type, compression-absorption-type. A prototype of ammonia-water compression-absorption heat pump is presented and modeled. (J.S.)

This paper is a series of transparencies presenting a comparative study of the thermal performances of different types of refrigerating machineries: di-thermal with vapor compression, tri-thermal with moto-compressor, with ejector, with free piston, adsorption-type, resorption-type, absorption-type, compression-absorption-type. A prototype of ammonia-water compression-absorption heat pump is presented and modeled. (J.S.)

The TIBERE Model is a neutron leakage method based on B{sub 1} heterogeneous transport equation resolution. In this work, we have studied the influence of the reflection mode at the boundary of the assembly. In particular the White boundary condition has been implemented in the APOLLO2 neutron transport code. We have compared the two TIBERE kinds of boundary conditions (specular and white) with the classical B{sub 1} homogeneous leakage method in the modelling of some reactors. We have remarked the better capability of the TIBERE Model to compute voided assemblies. The white boundary condition is also able to compute a completely voided assembly and, besides, wins a factor 10 in CPU time in comparison with the specular boundary condition. These two heterogenous leakage formalisms have been tested on a partially voided experiment and have shown that the TIBERE Model can compute this kind of situation with a greater precision than the classical B{sub 1} homogeneous leakage method, and with a shorter computational time. (author)

As nuclear orders are picking up a little, there are strengths competing against one another in the world industry of reactors, an industry that has been deeply affected for twenty years, by the smallness of the market and the reorganization of the electromechanical industry. Competition remains particularly difficult, even though, in terms of exports, national markets in industrialized countries such as the American market and European market are now open to foreign newcomers. One of the reasons of the difficulty is the increased commercial competition based on advanced reactor techniques untested due to strong faith in technology leading to forget the learning difficulties of older reactor types. On a narrow market, demanding and with very specific political interference, the reasoning is not like on an ordinary capital equipment market. Each builder tries to sell by relying on the assets it has in addition to the offered price and related services: industrial reputation and experience that play confusedly when untested advanced reactors are competing with one another, credit terms offered by the State and the government's influence on the market of emerging economies, the backing o the State's financial insurance in the event of risks taken in the sale of turnkey untested reactors. In the competition of the five manufacturers in the export market, American builders do not seem to have the best place, though even the leading position of Framatome ANP shows some limits. (author)

This day, organized by the SFEN, took place at Paris the 12 march 1998. Nine papers were presented. They take stock on the hybrid systems and more specifically the under-critical reactors. One of the major current preoccupation of nuclear industry is the problems of the increase of radioactive wastes produced in the plants and the destruction of the present stocks. To solve these problems a solution is the utilisation of hybrid systems: the coupling of a particle acceleration to an under-critical reactor. Historical aspects, advantages and performances of such hybrid reactors are presented in general papers. More technical papers are devoted to the spallation, the MUSE and the TARC experiments. (A.L.B.)

The transmutation of nuclear wastes in accelerator-driven nuclear reactorsoffers undeniable advantages. But before going into the detailed study of a particular project, we should (i) examine the possible applications of such systems and (ii) compare the different configurations, in order to guide technological decisions. We propose an approach, answering both concerns, based on the complete description of hybrid reactors. It is possible, with only the transmutation objective and a few technological constraints chosen a posteriori, to determine precisely the essential parameters of such reactors: number of reactors, beam current, size of the core, sub-criticality... The approach also clearly pinpoints the strategic decisions, for which the scientist or engineer is not competent. This global scheme is applied to three distinct nuclear cycles: incineration of solid fuel without recycling, incineration of liquid fuel without recycling and incineration of liquid fuel with on-line recycling; and for two spectra, either thermal or fast. We show that the radiotoxicity reduction with a solid fuel is significant only with a fast spectrum, but the incineration times range from 20 to 30 years. The liquid fuel is appropriate only with on-line recycling, at equilibrium. The gain on the radiotoxicity can be considerable and we describe a number of such systems. The potential of ADS for the transmutation of nuclear wastes is confirmed, but we should continue the description of specific systems obtained through this approach. (author)

The awareness of the international community and the convergence of scientific data about the global warming confirm the urgency of implementing greenhouse gases abatement technologies at the world scale. The growth of world energy demand will not allow to rapidly get rid of the use of fossil fuels which are the main sources of greenhouse gases. Therefore, the capture and disposal of CO{sub 2} is a promising way to conciliate the use of fossil fuels and the abatement of pollutants responsible for the global warming. The economical and industrial stakes of this technique are enormous. In front of the success of a first international colloquium on this topic held in Paris in 2005, the IFP, the BRGM and the Ademe have jointly organized a second colloquium in October 2007, in particular to present the first experience feedbacks of several pilot experiments all over the world. This document gathers the transparencies of 27 presentations given at this colloquium and dealing with: the 4. IPCC report on the stakes of CO{sub 2} capture and storage; the factor 4: how to organize the French economy transition from now to 2050; the technology perspectives, scenarios and strategies up to 2050; the European technological platform on 'zero-emission thermal plants'; the CO{sub 2} capture and storage road-map in the USA; research, development and implementation of CO{sub 2} capture and storage in Australia; the Canadian experience; ten years of CO{sub 2} capture and storage in Norway; the In Salah operations (Algeria); CO{sub 2} capture and storage: from vision to realisation; the oxi-combustion and storage pilot unit of Lacq (France); the Altmark gas field (Germany): analysis of CO{sub 2} capture and storage potentialities in the framework of a gas assisted recovery project; oil assisted recovery and CO{sub 2} related storage activities in Brazil: the Buracica and Miranga fields experience; carbon capture and storage, an option for coal power generation; steel-making industries and their CO{sub 2} capture and storage needs: the ULCOS program; CO{sub 2} capture technologies: road-maps and potential cost abatement; membranes: oxygen production and hydrogen separation; CO2GeoNet: integration of European research for the establishment of confidence in CO{sub 2} geologic storage; CO2SINK, CO{sub 2} geologic storage test at the European pilot site of Ketzin (Germany); storage in aquifers for European industrial projects: AQUA CO2; the US approach: US standards for the qualification of a CO{sub 2} storage in agreement with federal and state regulations; legal and regulatory aspects; societal acceptation; CO{sub 2} capture, geologic storage and carbon market; economic aspects of CO{sub 2} capture and storage; an experience of implementation of 'clean development mechanisms' in an industrial strategy; closing talk. (J.S.)

The present thesis deals with measures to optimize the large-volume production of Solid Oxide Fuel Cells (SOFC) based on thermal spraying technology. Based on the well-established Vacuum Plasma Spraying (VPS) at DLR the potential of alternative thermal spraying techniques as well as alternative base materials was investigated in order to deposit SOFC-anode, electrolyte and insulating layers. Production costs, reproducibility and long-term stability of the production process as well as the fuel cell performance were major target criteria. Depending on the parameter set applied when using the cost efficient Atmospheric Plasma Spraying (APS) in combination with Nickel-Graphite as base material a significant improvement of gas permeability and electrical conductivity was achieved in comparison to the VPS sprayed reference anode. The power density of a fuel cell with an APS-Nickel-Graphite anode (184 mW/cm{sup 2}) was slightly better than the performance with a VPS reference anode (159 mW/cm{sup 2}). In comparison to the VPS process, ceramic electrolyte layers of fully stabilized Zirconia (YSZ) with significantly higher gas tightness could be demonstrated when high energy processes such as Low Pressure Plasma Spraying (LPPS). Thin-film Low Pressure Plasma Spraying (LPPS-Thin-film) and High Velocity Oxy Fuel Spraying (HVOF) were applied. The power density of a fuel cell equipped with an HVOF electrolyte was significantly improved to 234 mW/cm{sup 2} as compared to 187 mW/cm{sup 2} with the VPS sprayed reference cell. Further improvement of the power density was achieved with an LPPS-electrolyte (273 mW/cm{sup 2}). HVOF and VPS sprayed layers of pure Spinel in composite with metallic active braze (equivalent to the sealing between individual layers in the fuel cell stack) could exceed the demanded charge transfer resistance of >1 k{omega}cm{sup 2} at 800 C operating temperature only in few cases. When blended base powder of Spinel and Magnesia in combination with the VPS was applied significantly higher charge transfer resistance of up to 1,7 M{omega}cm{sup 2} could be demonstrated at layer thicknesses between 20 {mu}m and 117 {mu}m. In the further course of the thesis an online controller was developed for the VPS process based on insitu process diagnostic systems in order to reproducibly produce thermal sprayed functional layers. In pretests the process diagnostic system Accuraspray appeared to be cable of identifying changes in the process by detecting particle properties in the spray torch. In order to serve as a knowledge base for the controller the plasma spraying process was successfully modelled as correlation between machine parameters, spray torch- und layer properties by using artificial neural networks (ANN). The ability for the generalization as well as the accuracy of the model was improved further by measures such as active learning (additional experiments at poorly represented regions) and the combination of several models with different network architectures as an ensemble (mixture of experts). Based on the artificial neural network model, an online process controller was developed, which manipulates major process parameters in case of a deviation in order to get back to the initial state. The ability of the prototype controller could be demonstrated successfully in most cases. (orig.)

The ASCOUF software has been developed par EDF to facilitate the quick analysis of defects contained in pipes or elbows of the primary loop. This preprocessing tool of Code Aster, the structural analysis finite element code of EDF, has been used to carry out, with an increase of productivity a series of numerical studies proving the mechanical strength of these components. Its validation, taking into account the feed-back from previous studies, leads us to rely on the results. ASCOUF has afterwards been extended to solve the problems of lack of thickness of pipes of the secondary loop. (authors)

In this thesis, rigorous steady and dynamic models have been developed for an industrial propylene polymerization process. They have then been used to develop new processes based on the existing one. For that purpose, this thesis has: ? reviewed the state of the art of the olefin Ziegler-Natta catalysts; ? analyzed classical propylene polymerization processes and their characteristics; ? analyzed the characteristics of industrial slurry and gas phase propylene polymerization reactors using re...

This thesis is interested in the political construction of environmental conflicts related to waste, industrial hazards and pollution management. The research is wondering about the political uses of danger, this is to say, about the devices the powerful, politicians, multinational companies, state representatives or NGOs leaders, used to fight against the threats, also produce and create them, make the threats growing bigger by lack of decision, or mask and exploit them. In order to find som...

Charbonnages de France, the French national coal board, has planned to stop its mining activities. Todays, its main goal concerns the remediation of its polluted mining and industrial sites. This article presents the cleansing techniques used at the Auby coking plant site for the removal of the polycyclic aromatic hydrocarbons (PAH) from the soil: protection of the aquifer, thermal desorption of the polluted earth, oxidation of the evaporated pollutants, valorization of the processed earth. (J.S.)

Full Text Available Hazard Mapping of Industrial Oil Storage in Hussein Dey, Algiers ( Algeria. The industrial hazardsare a danger to environment. It is the case of the fuel depot of Caroubier in the municpality of Hussein Dey in the outskirts of the capital Algiers. It is a potential danger in this urban area and a real challenge for the local authorities which do not have app ropriate tools for risk management. The municipality possesses single maps with colored items corresponding to the identification of spatial risks. The objective of this article is to map the risks generated by fuel depot. This mapping will help us localiz e the spatial vulnerability to suggest land planning measures.

The awareness of the international community and the convergence of scientific data about the global warming confirm the urgency of implementing greenhouse gases abatement technologies at the world scale. The growth of world energy demand will not allow to rapidly get rid of the use of fossil fuels which are the main sources of greenhouse gases. Therefore, the capture and disposal of CO{sub 2} is a promising way to conciliate the use of fossil fuels and the abatement of pollutants responsible for the global warming. The economical and industrial stakes of this technique are enormous. In front of the success of a first international colloquium on this topic held in Paris in 2005, the IFP, the BRGM and the Ademe have jointly organized a second colloquium in October 2007, in particular to present the first experience feedbacks of several pilot experiments all over the world. This document gathers the transparencies of 27 presentations given at this colloquium and dealing with: the 4. IPCC report on the stakes of CO{sub 2} capture and storage; the factor 4: how to organize the French economy transition from now to 2050; the technology perspectives, scenarios and strategies up to 2050; the European technological platform on 'zero-emission thermal plants'; the CO{sub 2} capture and storage road-map in the USA; research, development and implementation of CO{sub 2} capture and storage in Australia; the Canadian experience; ten years of CO{sub 2} capture and storage in Norway; the In Salah operations (Algeria); CO{sub 2} capture and storage: from vision to realisation; the oxi-combustion and storage pilot unit of Lacq (France); the Altmark gas field (Germany): analysis of CO{sub 2} capture and storage potentialities in the framework of a gas assisted recovery project; oil assisted recovery and CO{sub 2} related storage activities in Brazil: the Buracica and Miranga fields experience; carbon capture and storage, an option for coal power generation; steel-making industries and their CO{sub 2} capture and storage needs: the ULCOS program; CO{sub 2} capture technologies: road-maps and potential cost abatement; membranes: oxygen production and hydrogen separation; CO2GeoNet: integration of European research for the establishment of confidence in CO{sub 2} geologic storage; CO2SINK, CO{sub 2} geologic storage test at the European pilot site of Ketzin (Germany); storage in aquifers for European industrial projects: AQUA CO2; the US approach: US standards for the qualification of a CO{sub 2} storage in agreement with federal and state regulations; legal and regulatory aspects; societal acceptation; CO{sub 2} capture, geologic storage and carbon market; economic aspects of CO{sub 2} capture and storage; an experience of implementation of 'clean development mechanisms' in an industrial strategy; closing talk. (J.S.)

This report presents the results of a research work about 'the stakes of the dialogue around the follow up of nuclear and non nuclear industrial installations'. It used the experience of the North Cotentin radioecology group where expertise has been implemented in order to evaluate the impact on health of the releases of the Cogema La Hague plant. This report is the fruit of an interdisciplinary group ( experts of activities with risks, radiation protection, regulation in environment). (N.C.)

The production of wood pellets is a relatively new branch of the Swedish wood industry and has increased during the last years. A pilot study was performed to investigate the prevalence of airway symptoms, lung function and exposure among all 39 men employed in industrial production of wood pellets at six companies. The study included a questionnaire, medical examination, registration of nasal-PEF (peak expiratory flow) during a week, allergy screening (Phadiatop) and lung function (spirometry) before and after work shift. The results were compared with different reference data from other Swedish studies. Exposure measurements of monoterpenes and wood dust on filter and with a data logger (DataRAM) were also performed. The study group reported a higher frequency of cough without phlegm, awakening due to breathlessness and current asthma medication compared with reference data. For five of the six participants with physician-diagnosed asthma the disease debuted before the current employment and the results did not indicate an unusual asthma morbidity. Spirometry showed lower lung function before work shift than expected. However no difference over work shift was observed. A negative and non-significant correlation was seen between time with current work task and lung function. The study group reported a higher frequency of nasal symptoms mostly blockage, sneezing and dryness compared with reference data. The registrations of nasal-PEF did not show any differences between work and spare time. The prevalence of positive Phadiatop (23 %) did not differ from reference data. No association between exposure (wood dust and monoterpenes) and acute effects on lung function was observed. The wood dust exposure (0.16-19 mg/m{sup 3}) was high and 11 of 24 measurements exceeded the present Swedish occupational exposure limit of 2 mg/m{sup 3}. Peak exposures could be identified, e.g. at cleaning of engines with compressed air, with the DataRAM. The exposure to monoterpenes (0.64 and 24 mg/m{sup 3}) was low compared with the present Swedish limit of 150 mg/m{sup 3}. The monoterpene exposure does not seem to be a health or exposure problem in industrial production of wood pellets but wood dust exposure can effect the airways negatively. In this study the levels of wood dust were high and the study group reported more airway symptoms than expected. However the effect on lung function was small, but steps to reduce wood dust exposure should be done.

The aim of the policy of the French national gas utility, Gaz de France, towards industries, is to promote natural gas as a competitive fuel compared to fuels and electric power, with energy efficient solutions that are easy to implement and maintain: space heating, paint curing, surface cleaning, bath heating, vapour generation, waste treatment (especially for molding sand and volatile organic compounds, sludge drying). Gaz de France proposes also expertise schemes and audits

This report takes stock of the work carried out by the commission appointed by the French ministry of economy, finances and industry about the industrial and financial project of Electricite de France (EdF) in the framework of the liberalization of European energy markets. The report presents the conclusions of the commission about EdF's position in the new competition context, about the financial position of the group and about the foreseeable strategic options and their consequences in terms of equity fund needs. 5 appendixes present: the evolution of electricity prices, EdF and the energy policy, the electricity market and the competition in Europe, the EdF group: presentation and main adaptation stakes, the financial situation of EdF group. (J.S.)

This thesis presents a methodological approach to the design of computerized assistance for operators in control industrial processes. We are particularly interested in how to find the solutions which best suit their needs. Our preferred approach is focused on the operator: the main factors influencing his performance are reviewed and we make a synthesis which consists of a categorized list, or typology, of the extents of the operators` activities, tasks and errors. This typology is then used to classify the possible improvements as well as associated computer aids. The DIAPASON held system for fault diagnosis is integrated in this structure. This typology is our chosen basis for defining a specification method which enables the quality of the designed system to be guarantee. We propose a phased approach, the first phase of which involves analysing needs and thus identifying the objectives of the project. The second phase is the preparation of a performance specification which serves as a reference system for the project. In the third phase technical solutions are proposed to meet the requirements set out in the performance specification. The following phases involve studying the technical feasibility of the proposed solutions and the actual development of the system. Together with the feasibility study comes the step of making up a knowledge bank. The usual method of systems analysis are included in the typology of the aids. Furthermore, the SAGACE method uses a new approach to systems analysis based on its description which unites various points of view ; the evaluation of its possibilities forms a part of the construction of a reference system which gathers up the information needed to put the DIAPASON diagnosis system into action. (author).

At the request of the Ministries of Health and the Environment, IPSN is preparing and publishing the first version of the methodological guide devoted to managing industrial sites potentially contaminated by radioactive substances. This guide describes a procedure for defining and choosing strategies for rehabilitating such industrial sites. (author)

This conference is devoted to the radiation protection and the best way to optimize it. It reviews each area of the nuclear industry, and explores also the medical sector. Dosimetry, ALARA principle and new regulation are important points of this meeting. (N.C.)

This conference treats several subjects relative to the use of radioactive sources. The first session comprises three articles about ionizing sources and regulation. The second session, with three articles, tackles the question of radiation protection in the use of sources in industrial field. The third session, four articles, treats the same question but in the medicine and university media. The fourth session (three articles) is devoted to the organisation of radiation protection in the case of accidents. The fifth session concerns the management of spent sources (three articles). The sixth session studies the radiation protection of sources in Europe. The seventh and final session ends with the part and coordination of actors in radiation protection in the sources management (three articles). (N.C.)

The effects of power interruption on individual industrial customers by computing an individual nuisance index (IGI) is one of the tools planned by Hydro-Quebec to measure quality of service to its customers. When fully functioning, IGI will represent a combined value of loss of sales, overtime, lost materials and other direct costs, each IGI tailor-made for a particular company. Data for computing the index will be obtained from Hydro-Quebec`s own customer classification database, plus a commercial technical database (DTC) containing data required for the assessment of the nuisance, and a survey carried out by Hydro-Quebec involving some 1600 industrial customers. As of this date, the DTC is not yet available. A statistical analysis of survey responses was substituted to provide default values based on available parameters. Hydro-Quebec is confident that this new approach to evaluating service quality will open new horizons in quality assurance. 3 refs., 5 tabs.

The purpose of this study to valorize solid wastes from the Solvay process for the production of sodium carbonate as reagents for the treatment of H2S in air at concentrations typically found in wastewater treatment plants (tens to hundreds of ppmv of H2S). Firstly, the reactivity of two residues was evaluated in a gas-liquid-solid reactor at laboratory scale (250 ml). This reactor operates semi-continuously (continuous passage of gas through a fixed volume of slurry). The influence of operat...

This paper describes the French Promill Company activity in the design and manufacturing of complete drying-crushing-agglomerating units for agro-industrial products (pulp of beet, lucerne, etc..). The paper focusses on the thermal and mechanical efficiency of the high temperature dryer and on the pulp granulating squeezer. (J.S.)

Power frequency fields are electric and magnetic fields produced by electric power. The first evidence to show that power frequency fields may have an effect on human health was noted in 1972 when it was reported that workers in high voltage switchyards suffered from a number of ailments. Fundamentals of electric and magnetic fields, and methods of their measurement, are explained. Some sources of power frequency fields are reviewed, and measured occupational exposures to these fields by electric utility workers are listed. Possible biological and health effects from power frequency fields are summarized from epidemiological, whole-animal, cellular, and observational studies reported in the literature. These include increased risk of cancer, decrease in melatonin, reproductive and developmental abnormalities, decreased heart rates, changes in calcium flow across cell membranes, and changes in neurotransmitters and corticosterone. Exposure standards for power frequency fields are given and methods for controlling exposure to those fields are noted. 53 refs., 1 fig., 2 tabs.

The documentation management of a big industrial project involves a continuous update of information, both in the study and realization phase or in the operation phase. The organization of the technical documentation for big industrial projects requests complex information systems. In the first part of this paper are presented the methods appropriate for the analysis of documentation management procedures and in the second part are presented the tools by the combination of which a documentation system for the user is provided. The case of the documentation centres for the Hague reprocessing plant is described.

The French society Promill has developed complete units for the drying, crushing and aggregating of agro-industrial products (beet roots, agricultural wastes, lucerne, maize, etc.). Drying is conducted in a three-pass drum, using any type of fuel (fuel oil, gas, electric power, coal), and ensuring a thermal yield of 680 kCal/kg and ash emission rates complying with French and European legislation. Granulation is conducted with vapour addition, with a granulate flowrate reaching 15 T/h. Crushing is carried out in a hammer mill

This report presents the results of a research work about 'the stakes of the dialogue around the follow up of nuclear and non nuclear industrial installations'. It used the experience of the North Cotentin radioecology group where expertise has been implemented in order to evaluate the impact on health of the releases of the Cogema La Hague plant. This report is the fruit of an interdisciplinary group ( experts of activities with risks, radiation protection, regulation in environment). (N.C.)

The aim of this document is the description of the implementing mechanism concerning the wind energy, of February 2000 for the principles and July 2006 for the investment program of the Government on the purchase obligation of the wind electricity by EDF. The laws and orders, the cost of the wind energy are detailed and discussed. (A.L.B.)

Natural gas supplied to industry for process utilizations originates from different sources and that can cause fluctuations in gas composition. Changing gas composition may lead to production problems in industry with sensitive thermal processes (particularly glass industry and thermal metal treatments), such as efficiency and product quality. An equipment suitable to control and adjust such variations has been developed. Experimental tests in laboratory were carried out in order to investigate the control system accuracy and reliability. In particular five different settings were tested: at a preset thermal input by adjusting the natural gas flow rate in respect to Wobbe Index variations; at a set furnace temperature and stack oxygen level with variable thermal input by monitoring the Wobbe Index value; at constant Wobbe Index value by adding air to natural gas; at constant thermal input and prefixed Wobbe Index value by adding air to natural gas and varying the air and gas mixture flow rate; gross calorific value control by adding air or LPG to natural gas. All the tested settings gave good results. This report illustrates these results and the main features of the control system. The control and regulation system was installed in two glass factories for field tests. (authors)

Full Text Available This article presents the outcomes of a study on the practises of Gabonese teachers who teach in the fields of industrial technical and vocational education. This study is interested in their daily activities and aims at characterising the inevitable variation which exists between the knowledge with work in the production activities industrial in company and those implemented in the institutions of vocational training. More generally, it is a question of appreciating the relation between training and employment. Thus, we are interested in the orientation given by the teacher to his daily practise to reduce or not this variation. The results of this study highlight the impact of epistemological and pragmatic dimensions on the organisation of the activities and thus on the practises. Obviously, the level of entry in the technical school course, the first job, the transport conditions, the appropriation of the reference frame of training and the development of the contents of courses are some of the items which impact this process of reduction of the difference between the two forms of knowledge.

International audience; Through a variety of events and exhibitions, the concept of “sensory heritage” has appeared in several forms of expression of the cities’ industrial memory. It relies on the assumption that there would exist sensory aspects of a city related to its past or present activities, which would constitute its identity and therefore be subject to a particular conservation and diffusion. This paper will discuss the connection between these two apparently opposite notions. In a ...

This paper deals with the first results of some research into the field of Business-to-Business, involving some processes of subcontracting implemented at an international level by SMEs operating in a high tech sector (PCBs). Particular attention has been placed on the prospects for development of the SME, which are located in the industrial district of Castelfidardo, and which operate in the global market. This research has the aim of studying the modifications that have effected the demand ...

The practice shows that many crucial decisions for industrialization in developing countries have been taken based on incomplete information. For strategic decisions an incomplete information may have catastrophic consequences. The function of policy-making is defined as the process by which the information generated/or used in a particular context is reevaluated in a different context in order to formulate/or execute a policy of alternative decisions. It follows that the industrial information must be presented in such a manner to allow a reevaluation and alternative decisions. 30 notes.

Modern biotechnology was 'invented' in 1972/73. New technologies typically require a fifteen to twenty year incubation period of research and development before their commercial, industrial and other economic impacts begin to become significant. Biotechnology holds the key to many problems from environmental management to major savings for the consumer through improved performance of products and processes as well as from entirely new products and services. Simultaneously, environmental applications of biotechnology will produce enormous savings in the economic costs of pollution. Modern biotechnology, its current applications and its potential to reduce pressure on the environment, will be discussed in respect to biosafety, regulatory approaches, new processes, R and D, 'after processes' waste treatment and bioremediation. (orig.).

A 5 MW installation is described which is a reduced model of the heat exchange system of a sodium-cooled reactor. This plant, which is situated at Grand Quevilly (near Rouen), consists of: 1 - A primary sodium loop made up of a sodium re-heater running on heavy diesel oil, a mechanical pump and an intermediate exchanger made up of clusters of tubes fitted with baffles. 2 - A NaK(56 per cent of K) secondary loop consisting mainly of a mechanical pump and a double-wall steam generator with forced circulation and complete vaporization. 3 - A tertiary water loop consisting of the inside of the steam generator pipes, a pressure-reducing valve which cools down the super-heated fluid and acts as a turbine, a condenser, a charge-pump and a supply pump for the boiler. The heat is given finally to a water-source flowing into the Seine. Two important points of the installation are: - The water treatment unit - The control and regulation system Apart from the general satisfactory operation of the installation which it is hoped to obtain, a careful study will be made of the heat transmission coefficients of the important equipment such as the intermediate exchanger and the steam generator. The construction was finished on April 28, 1964. (author) [French] On decrit une installation de 5 MW figurant a echelle reduite un systeme de transfert de chaleur d'un reacteurrefroidi au sodium. Cette installation, situee a Grand Quevilly (pres de Rouen) comprend: 1 - Une boucle de sodium primaire comportant un rechauffeur de sodium alimente en fuel lourd, une pompe mecanique et un echangeur intermediaire a faisceau tubulaire muni de chicanes, 2 - Une boucle de NaK (56% de K) secondaire dont les appareils principaux sont une pompe mecanique et un generateur de vapeur a double paroi, circulation forcee et vaporisation totale. 3 - une boucle tertiaire a eau comprenant l'interieur des tubes du generateur de vapeur, un detendeur-desurchauffeur simulant une turbine, un condenseur, une pompe de

The aim of this document is the main improvements of the CANDU reactors in the economic, safety and performance domains. The presentation proposes also other applications as the hydrogen production, the freshening of water sea and the bituminous sands exploitation. (A.L.B.)

This PhD dissertation focuses on the improvement of very high temperature thermometry (1100 deg. C to 2480 deg. C), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: - The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (±0.001 deg. C) to applied research with a reasonable degradation of uncertainties (±3-5 deg. C). - The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300 deg. C) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000 deg. C)

In this work, we looked for the error sources in the calculation of the isothermal temperature coefficient for light water lattices. We studied three fields implied: the nuclear data, the calculation methods and the temperature coefficient measurement. About the measurement, we pointed out the difficulties of he interpretation. So we used an indirect approach by the mean of critical states at various temperatures. In that way, we can say that if the errors in the effective multiplication factor are constants with temperature then the temperature coefficient is correctly calculated. We studied the neutronic influence of light water models which are used in the thermal scattering cross-section computation. This cross-section determines the thermalization process of neutrons. We showed that the actual model (JEF2) is satisfactory of the needs of the reactors physics. Concerning the majors isotopes ({sup 235}U, {sup 238}U, {sup 239}Pu), the uncertainties on the nuclear data do not seem as a preponderant cause of errors, without to be totally negligible. We also studied, with the neutron transport code Apollo-2, the influence of difference approximations for cell calculation . The new possibilities of the code has been used to represent the critical experiments, particularly the improvement of the resonance self-shielding formalism. The calculation scheme adopted permits to remove partially the fundamental mode approximation by the mean of a two-dimensional transport calculation with the SN method, the axial leakage being treated as an absorption in DB{sup 2}{sub Z}. The agreement between theory and experiment is good both for the reactivity and the temperature coefficient. (author). 114 refs., 40 figs., 163 tabs., 1 append.

The fuels and targets program for transmutation, performed in the framework of the axis 1 of the December 1991 law about the researches on the management of long-lived radioactive wastes, is in perfect consistency with the transmutation scenario studies carried out in the same framework. These studies put forward the advantage of fast breeder reactors (FBR) in the incineration of minor actinides and long-lived fission products. The program includes exploratory and technological demonstration studies covering the different design options. It aims at enhancing our knowledge of the behaviour of materials under irradiation and at ensuring the mastery of processes. The goals of the different experiments foreseen at Phenix reactor are presented. The main goal is to supply a set of results allowing to precise the conditions of the technical feasibility of minor actinides and long-lived fission products incineration in FBRs. (J.S.)

During the core loading period of a PWR, any fuel assembly misplacements may significantly reduce the existing criticality margin. The Dampierre 4-18 event showed the present monitoring based on the variations of the outside-core detector counting rate cannot detect such misplacements. In order to circumvent that, a more detailed analysis of the available signal was done. We particularly focused on the neutronic noise analysis methods such as MSM (modified source multiplication), MSA (amplified source multiplication), Rossi-{alpha} and Feynman-{alpha} methods. The experimental part of our work was dedicated to the application of those methods to a research reactor. Finally, our results showed that those methods cannot be used with the present PWR instrumentation. Various detector positions were then studied using Monte Carlo calculations capable of following the neutron origin. Our results showed that the present technology does not allow us to use any solution based on neutron detection for monitoring core loading. (author)

The complete study of molten salt reactors, designed for a massive and durable nuclear energy production, must include neutronics, hydraulics and thermal effects. This coupled study, using the MCNP and Trio{sub U} codes, is undertaken in the case of the MSRE (molten salt reactor experiment) prototype. The obtained results fit very well the experiment. Their extrapolation suggests ways of improving the safety coefficients of power molten salt reactors. A second part is devoted to accelerator driven subcritical reactors, developed to incinerate radioactive waste.We propose a method to measure the prompt reactivity from the decay following a neutron pulse. It relies only on the distribution of times between generations, which is a characteristic of the reactor. This method is implemented on the results of the MUSE 4 experiment, and the obtained reactivity is accurate within 5%. (author)

The conversion of natural gas to synthesis gas (mixture of H{sub 2} and CO) is a main challenge for the hydrogen and clean fuels production. Mixed (ionic O{sup 2-} and electronic) conducing ceramics membrane reactors seem particularly promising. The design considered for the membrane is a tri-layer system integrating a reforming catalyst and a dense membrane laying on a porous support. Among the materials considered for the dense membrane, perovskites La{sub 1-x}Sr{sub x}Fe{sub 1-y}Ga{sub y}O{sub 3-{delta}} seem to be interesting for their performances and stability. The oxygen flux through the membrane is measured in terms of temperature under different oxygen partial pressure gradients. In the industrial experimental conditions, the membrane is submitted to a strong oxygen (air/methane) partial pressure gradient of about 900 C which induces mechanical stresses, on account of the material expansion difference, in terms of p{sub O2}. In this framework, the evolutions of the performances and of the expansion coefficient have been followed in terms of the substitutions rates in La{sub (1-x)}Sr{sub x}Fe{sub (1-y)}Ga{sub y}O{sub 3-{delta}} with x{<=}0.5 and y{<=}0.5. (O.M.)

The main goal of the Borexino experiment is to measure in real time the solar neutrino flux from the beryllium (Be{sup 7}) line at 862 keV. Beyond this pioneer low energy neutrino detection, Borexino will be able to measure solar neutrinos above the MeV, (B{sup 8} neutrinos and pep neutrinos), nuclear reactor neutrinos (with an average energy of 3 MeV) and the supernova neutrinos (their spectrum goes up to some ten MeV). In this work I mainly focus on the study of the nuclear reactors neutrinos. This field has recently been enriched by the results of the KamLAND experiment, which have greatly improved the determination of the neutrino oscillation parameters. In order to measure these events which are above the MeV, the Borexino collaboration entrusted the PCC group at College de France, with the tasks of developing a fast digit system running at 400 MHz: the FADC cards. The PCC group designed the FADC cards and completed them at the beginning of 2002. The first cards which were introduced in the main electronic acquisition unit allowed us to control their functioning and that of the acquisition software. FADC cards were also installed in the Borexino prototype, CTF. The data are analysed in order to determine a limit to the expected background noise of Borexino in measuring the nuclear reactor neutrinos. (author)

With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

The product software CIVACUVE is used by INTERCONTROLE for the analysis of UT examinations, for detection, performed by the In-Service Inspection Machine (MIS) of the vessels of nuclear power plants. This software is based on an adaptation of an algorithm of SEGMENTATION (CEA CEREM), which is applied prior to any analysis. It is equipped with tools adapted to industrial use. It allows to: - perform image analysis thanks to advanced graphic tools (Zooms, True Bscan, 'contour' selection...), - backup of all data in a database (complete and transparent backup of all informations used and obtained during the different analysis operations), - connect PC to the Database (export of Reports and even of segmented points), - issue Examination Reports, Operating Condition Sheets, Sizing curves... - and last, perform a graphic and numerical comparison between different inspections of the same vessel. Used in Belgium and France on different kind of reactor vessels, CIVACUVE has allowed to show that the principle of SEGMENTATION can be adapted to detection exams. The use of CIVACUVE generates a important time gain as well as the betterment of quality in analysis. Wide data opening toward PC's allows a real flexibility with regard to client's requirements and preoccupations.

During an experiment on the zero power heavy water reactor with natural uranium fuel in the Boris Kidric Institute of Nuclear Sciences, the reactor escaped control. Six staff members in the immediate surrounding of the bare assembly were exposed to high neutron and ionising irradiation. Other two employees who were at some bigger distance were exposed to doses higher than permitted. This paper deals with the circumstances that caused the accident, status of the dosimetry, control and alarm systems. Individual exposure doses were estimated according to the calculated neutron flux values obtained from measuring the activities of personal belongings made of gold and copper as well as radioactive phosphorous from urine.

The fast ultrasonic visualization under sodium is in the programme of research and development on the inspection inside the fast neutron reactors. This work is about the development of a such system of fast ultrasonic imaging under sodium, in order to improve the existing visualization systems. This system is based on the principle of orthogonal imaging, it uses two linear antennas with an important dephasing having 128 piezo-composite elements of central frequency equal to 1.6 MHz. (N.C.)

Industrial oxalic precipitation processed in an un-baffled magnetically stirred tank, the Vortex Reactor, has been studied with uranium simulating plutonium. Modelling precipitation requires a mixing model for the continuous liquid phase and the solution of population balance for the dispersed solid phase. Being chemical reaction influenced by the degree of mixing at molecular scale, that commercial CFD code does not resolve, a sub-grid scale model has been introduced: the finite mode probability density functions, and coupled with a model for the liquid energy spectrum. Evolution of the dispersed phase has been resolved by the quadrature method of moments, first used here with experimental nucleation and growth kinetics, and an aggregation kernel based on local shear rate. The promising abilities of this local approach, without any fitting constant, are strengthened by the similarity between experimental results and simulations. (author)

The Service of the Reactor Siloe (CEA/DRN/DRE/SRS), fully aware of the abilities and knowledge of his teams in the field of research reactor operating, has undertaken a project of knowledge engineering in this domain. The following aims have been defined: knowledge capitalization for the installation in order to insure its perenniality and valorization, elaboration of a project for the aid of the reactor operators. This article deals with the different actions by the SRS to reach the aims: realization of a technical model for the operation of the Siloe reactor, development of a knowledge-based system for the aid for operating. These actions based on a knowledge engineering methodology, SAGACE, and using industrial tools will lead to an amelioration of the security and the operating of the Siloe reactor. (authors). 13 refs., 7 figs.

The neutronic calculation of a reactor's core is usually done in two steps. After solving the neutron transport equation over an elementary domain of the core, a set of parameters, namely macroscopic cross sections and potentially diffusion coefficients, are defined in order to perform a full core calculation. In the first step, the cell or assembly is calculated using the "fundamental mode theory", the pattern being inserted in an infinite lattice of periodic structures. This simple representation allows a precise modeling for the geometry and the energy variable and can be treated within transport theory with minimalist approximations. However, it supposes that the reactor's core can be treated as a periodic lattice of elementary domains, which is already a big hypothesis, and cannot, at first sight, take into account neutron leakage between two different zones and out of the core. The leakage models propose to correct the transport equation with an additional leakage term in order to represent this phenomenon. For historical reasons, numerical methods for solving the transport equation being limited by computer's features (processor speeds and memory sizes), the leakage term is, in most cases, modeled by a homogeneous and isotropic probability within a "homogeneous leakage model". Driven by technological innovation in the computer science field, "heterogeneous leakage models" have been developed and implemented in several neutron transport calculation codes. This work focuses on a study of some of those models, including the TIBERE model from the DRAGON-3 code developed at Ecole Polytechnique de Montreal, as well as the heterogeneous model from the APOLLO-3 code developed at Commissariat a l'Energie Atomique et aux energies alternatives. The research based on sodium cooled fast reactors and light water reactors has allowed us to demonstrate the interest of those models compared to a homogeneous leakage model. In particular, it has been shown that a heterogeneous model has a significant impact on the calculation of the out of core leakage rate that permits a better estimation of the transport equation eigenvalue Keff . The neutron streaming between two zones of different compositions was also proven to be better calculated.

This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

The analysis of DOEL-1 in-vessel and ex-vessel neutron dosimetry, using the DOT 3.5 Sn code coupled with the VITAMIN-C cross-section library, showed the same C/E values for different detectors at the surveillance capsule and the ex-vessel cavity positions. These results seem to be in contradiction with those obtained in several Benchmark experiments (PCA, PSF, VENUS...) when using the same computational tools. Indeed a strong decreasing radial trend of the C/E was observed, partly explained by the overestimation of the iron inelastic scattering. The flat trend seen in DOEL-1 could be explained by compensating errors in the calculation such as the backscattering due to the concrete walls outside the cavity. The `Concrete Benchmark` experiment has been designed to judge the ability of this calculation methods to treat the backscattering. This paper describes the `Concrete Benchmark` experiment, the measured and computed neutron dosimetry results and their comparison. This preliminary analysis seems to indicate an overestimation of the backscattering effect in the calculations. (authors). 5 figs., 1 tab., 7 refs.

This report focuses on the current knowledge, updated at the end of 1999, about the physicochemical state of the fuels leaving light water reactors, and particularly pressurized water reactors. Lessons are withdrawn from it making it possible to determine the points which require a necessary deepening of the data and coherence of interpretations. Lastly, evolution of the sailed fuel rod as well as the potential availability of gases and volatile fission products, during a secular storage or of a multi-millennium disposal, are the subject of an attempt at forecast. Accessible data in the scientific literature, or those acquired at the CEA, are particularly numerous. Their analysis and their synthesis are joined together to constitute a collection of references intended to the specialists in nuclear fuel and for all those which contribute to the reflexion on the storage or final disposal of the irradiated fuel. This memory is structured in ten chapters. The last chapter makes it possible to retain on some pages, the essential lessons of this study. Chapter I: Introduction; Chapter II: Characteristics of assemblies and fuels before irradiation; Chapter III: Transformations in reactor; Chapter IV: State of rods leaving the reactor; Chapter V: State of pellets; Chapter VI: Chemical and structural composition of the fuel; Chapter VII: Fuel fragmentation and density; Chapter VIII: Phenomena at the pellet periphery. Formation, characteristics and structure of the rim.Chemical interaction between pellet and cladding; Chapter IX: Location of fission gases and volatile fission products; Chapter X: Review, lessons and predictions. (authors)

The transmutation of nuclear wastes in accelerator-driven nuclear reactorsoffers undeniable advantages. But before going into the detailed study of a particular project, we should (i) examine the possible applications of such systems and (ii) compare the different configurations, in order to guide technological decisions. We propose an approach, answering both concerns, based on the complete description of hybrid reactors. It is possible, with only the transmutation objective and a few technological constraints chosen a posteriori, to determine precisely the essential parameters of such reactors: number of reactors, beam current, size of the core, sub-criticality... The approach also clearly pinpoints the strategic decisions, for which the scientist or engineer is not competent. This global scheme is applied to three distinct nuclear cycles: incineration of solid fuel without recycling, incineration of liquid fuel without recycling and incineration of liquid fuel with on-line recycling; and for two spectra, either thermal or fast. We show that the radiotoxicity reduction with a solid fuel is significant only with a fast spectrum, but the incineration times range from 20 to 30 years. The liquid fuel is appropriate only with on-line recycling, at equilibrium. The gain on the radiotoxicity can be considerable and we describe a number of such systems. The potential of ADS for the transmutation of nuclear wastes is confirmed, but we should continue the description of specific systems obtained through this approach. (author)

The aim of this report is to find a method for estimating the anti-reactivity of control rods perpendicular to the axis in a cylindrical pile. The paper is divided into two parts. In the first is given a method of calculating control rods in a rectangular pile, similar to the Nordheim-Scalettar method for cylindrical piles. As an example the formulas are given for the theories of one and two neutron groups, the generalisation for several groups being evident. In the second part we find by a variation method a formula for estimating the Laplacian of a pile, which may be divided into parallelepipeds for which the Laplacian are given. Finally, this formula is used to calculate the anti-reactivity of rods perpendicular to the axis in a cylindrical pile. (author) [French] Le but de ce rapport est de trouver une methode pour estimer l'antireactivite des barres de controle perpendiculaires a l'axe dans pile cylindrique. Le rapport se divise en deux parties. Dans la premiere nous donnons une methode de calcul des barres de controle dans une pile rectangulaire, analogue a la methode de Nordheim-Scalettar pour les piles cylindriques. A titre d'exemple, nous donnons les formules de theories a un et deux groupes de neutrons, la generalisation pour plusieurs groupes est evidente. Dans la deuxieme partie, nous trouvons, par une methode de variation, une formule qui permet d'estimer le laplacien d'une pile, qui peut etre divisee en parallelepipedes dont les laplaciens sont donnes. Nous utilisons enfin, cette formule pour calculer l'antireactivite des barres perpendiculaires a l'axe dans une pile cylindrique. (auteur)

The author aims to present the little-known story of the Oklo natural reactors. He recalls the historical aspects of the Oklo reactors discovery by the CEA in 1972, he explains the scientific phenomenon and the interest, notably as a 'natural analogue' for the geological disposal of high level radioactive wastes. (A.L.B.)

This document makes a detailed technical evaluation of three families of separation-transmutation scenarios for the management of radioactive wastes. These scenarios are based on 2 parks of reactors which recycle plutonium and minor actinides in an homogeneous way. A first scenario considers the multi-recycling of Pu and Np and the mono-recycling of Am and Cm using both PWRs and FBRs. A second scenario is based on PWRs only, while a third one considers FBRs only. The mixed PWR+FBR scenario requires innovative options and gathers more technical difficulties due to the americium and curium management in a minimum flux of materials. A particular attention has been given to the different steps of the fuel cycle (fuels and targets fabrication, burnup, spent fuel processing, targets management). The feasibility of scenarios of homogeneous actinides recycling in PWRs-only and in FBRs-only has been evaluated according to the results of the first scenario: fluxes of materials, spent fuel reprocessing by advanced separation, impact of the presence of actinides on PWRs and FBRs operation. The efficiency of the different scenarios on the abatement of wastes radio-toxicity is presented in conclusion. (J.S.)

In nuclear reactors, whatever the type considered, Pressurized Water Water Reactors (PWRs), Fast Breeder reactors (FBRs)..., thermal-hydraulics, the science of fluid mechanics and thermal behaviour, plays an essential role, both in nominal operating and accidental conditions. Fluid can either be the primary fluid (liquid or gas) or a very specific fluid called corium, which, in case of severe accident, could result from core and environning structure melting. The work reported here represents a 20-year contribution to thermal-hydraulic issues which could occur in FBRs and PWRs. Working on these two types of reactors, both in nominal and severe accident situations, has allowed me to compare the problems and to realize the importance of communication between research teams. The evolution in the complexity of studied problems, unavoidable in order to reduce costs and significantly improve safety, has led me from numerical modelling of single-phase flow turbulence to high temperature real melt experiments. The difficulties encountered in understanding the observed phenomena and in increasing experimental databases for computer code qualification have often entailed my participation in specific measurement device developments or adaptations, in particular non-intrusive devices generally based on optical techniques. Being concerned about the end-use of this research work, I actively participated in `in-situ` thermalhydraulic experiments in the FBRs: Phenix and Super-Phenix, of which I appreciated their undeniable scientific contribution. In my opinion, the thermal-hydraulic questions related to severe accidents are the most complex as they are at the cross-roads of several scientific specialities. Consequently, they require a multi-disciplinary approach and a continuous see-saw motion between experimentalists and modelling teams. After a brief description of the various problems encountered, the main ones are reported. Finally, the importance for research teams to participate in large international concerted actions is highlighted. (author) 64 refs.

Gas-solid reactions control a great number of major industrial processes involving matter transformation. This dissertation aims at showing that mathematical modelling is a useful tool for both understanding phenomena and optimising processes. First, the physical processes associated with a gas-solid reaction are presented in detail for a single particle, together with the corresponding available kinetic grain models. A second part is devoted to the modelling of multiparticle reactors. Different approaches, notably for coupling grain models and reactor models, are illustrated through various case studies: coal pyrolysis in a rotary kiln, production of uranium tetrafluoride in a moving bed furnace, on-grate incineration of municipal solid wastes, thermogravimetric apparatus, nuclear fuel making, steel-making electric arc furnace. (author)

A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

This document underlines the importance of water shortage in many areas in the world in the future. The water sea desalination can be a efficient solution to this problem. The desalination methods are presented. In this context the desalination reactors appear as a competitive solution, facing the fossil energies systems not only for the simultaneous electric power and drinking water production, but also for the minimization of greenhouse gases. (A.L.B.)

27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

This paper describes general features concerning development in the field of fast reactors in France from March 1980 to March 1981. It concentrates mainly on: Rapsodie, Phenix NPP, prototype reactor Super Phenix 1, future fast reactor NPPs and current research and development programs in the field. The present situation is as follows. Rapsodie has restarted operation but at reduced power in July 1980 because of the problems in the primary circuit which have not yet been solved. Phenic operates in a very satisfactory manner. Construction of Super Phenix is continuing normally. Research activities are performed sometimes for the needs of Super Phenix and sometimes for the needs of future fast rector projects like Super Phenix 2. International cooperation is being continued.

After 50 year long feedback experience on nuclear reactor operations it is legitimate to wonder whether experimental facilities used to support nuclear power programs are still necessary. The various participants of this conference said yes for mainly 4 reasons: -) to validate the extension of the service life of a reactor without putting at risk its high safety standard, -) to give the reactor more flexibility to cope with the power demand, -) to confront the results given by computerized simulations with experimental data, and -) to qualify the nuclear systems of tomorrow. (A.C.)

This work deals with modelling, simulation and optimization of the coupling between nuclear reactors (PWR, modular high temperature reactors) and desalination processes (multiple effect distillation, reverse osmosis). The reactors considered in this study are PWR (Pressurized Water Reactor) and GTMHR (Gas Turbine Modular Helium Reactor). The desalination processes retained are MED (Multi Effect Distillation) and SWRO (Sea Water Reverse Osmosis). A software tool: EXCELEES of thermodynamic modelling of coupled systems, based on the Engineering Algebraic Equation Solver has been developed. Models of energy conversion systems and of membrane desalination processes and distillation have been developed. Based on the first and second principles of thermodynamics, these models have allowed to determine the optimal running point of the coupled systems. The thermodynamic analysis has been completed by a first economic evaluation. Based on the use of the DEEP software of the IAEA, this evaluation has confirmed the interest to use these types of reactors for desalination. A modelling tool of thermal processes of desalination in dynamic condition has been developed too. This tool has been applied to the study of the dynamics of an existing plant and has given satisfying results. A first safety checking has been at last carried out. The transients able to jeopardize the integrated system have been identified. Several measures aiming at consolidate the safety have been proposed. (O.M.)

Boiling Water Reactors represent one third of the world's reactors. They are presently evolving towards greater simplification, allowing a reduction in the costs of operation, improved safety and a relative flexibility in their capacity to accommodate 100% MOX cores. The CEA, in a combined effort with its partners, the COGEMA and the EDF, would like to assess the interest of this reactor type, especially on this last point. A definition program and subsequent qualification of the calculation scheme have been undertaken. We are presenting here the specific features inherent in the calculation of these reactors, in comparison to PWRs, as well as the first results of the program. (authors)

{beta}{sub eff}, which represents the effective delayed neutron fraction, is an important parameter for the reactor nominal working as well as for studies of its behaviour in accidental situation. In order to improve the safety of nuclear reactors, we propose here to validate its calculation by using the ERANOS code with ERALIB1 library and by taking into account all the fission process physics through the {nu} energy dependence. To validate the quality of this calculation formalism, we calculated uncertainties as precisely as possible. The experimental values of {beta}{sub eff}, as well their uncertainties, have also been re-evaluated for consistency, because these `experimental` values actually contain a calculated component. We therefore obtained an entirely coherent set of calculated and measured {beta}{sub eff}. The comparative study of the calculated and measured values pointed out that the JEF2.2 {nu}{sub d} are already sufficient because the (E-C)/C are inferior to 3 % in average and in their uncertainly bars. The experimental uncertainties, even if lightly superior to those previously edited, remain inferior to the uncertainties of the calculated values. This allowed us to fit {nu}{sub d} with {beta}{sub eff}. This adjustment has brought an additional improvement on the recommendations of the {nu}{sub d} average values, for the classical scheme (thermal energy, fast energy) and for the new scheme which explains the {nu}{sub d} energy dependence. {beta}{sub eff}, for MOX or UOX fuel assemblies in thermal or fast configurations, can therefore be obtained with an uncertainty due to the nuclear data of about 2.0 %. (author) 110 refs.

The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

The aim of this technical meeting is the improvement of the existing nuclear data and the production of new data of interest for the upstream and downstream of the fuel cycle (enrichment, fabrication, management, storage, transport, reprocessing), for the industrial reactors, the research reactors and the new reactor concepts (criticality, dimensioning, exploitation), for the instrumentation systems (external and internal sensors), the radioprotection, the residual power, the structures (neutron bombardment effect on vessels, rods etc..), and for the activation of steel structures (Fr, Ni, Co). The expected result is the collection of more reliable and accurate data in a wider spectrum of energies and temperatures thanks to more precise computer codes and measurement techniques. This document brings together the communications presented at this meeting and dealing with: the process of production and validation of nuclear data; the measurement facilities and the big international programs; the users needs and the industrial priorities; the basic nuclear data (BND) needs at Cogema; the expression and evaluation of BND; the evaluation work: the efficient cross-sections; the processing of data and the creation of activation libraries; from the integral measurement to the qualification and the feedback on nuclear data. (J.S.)

Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

This study concerns the implementing and the exact determination of the gains of the plant control loops and the knowledge of the system response during possible solicitations under normal conditions. (A.L.B.)

Ward, A. [The Royal College of Science and Technology, Glasgow, Scotland (United Kingdom)

1963-03-15

The Scottish Research Reactor Centre is now under construction and will be completed and in operation by the summer of 1963. The reactor is 100 kW of the tank type with water cooling and water/graphite moderation using enriched U{sup 235} fuel. The experimental facilities include a large thermal column, a large shield experiment water tank and a radioisotope production facility with transfer rabbit tubes. There are effectively three through tubes in the central core; one through tube in the thermal column, several small central vertical stringers and one six-inch square vertical stringer penetrating to the centre of the core. Many horizontal stringers pass through the thermal column, the central one penetrating to within one inch of a fuel tank. The reactor facilities are supported by a wide variety of adjacent small laboratories. These include hot source handling and preparation facilities, changing rooms, electrical and mechanical workshops, darkrooms, microcurie laboratories, animal house, biological and chemical laboratories, low-background counting room, lecture theatre and a library. It is expected that the research will extend over many scientific and technological disciplines; a good proportion of the work will involve short-lived radioisotopes and typical projects are described. (author) [French] Le reacteur de recherche ecossais actuellement en construction sera termine et entrera en service pendant l'ete 1963. Il s'agit d'un reacteur-piscine de 100kW a uranium-235 enrichi, refroidi a l'eau eti ralenti a l'eau et au graphite. Les installations experimentales comprennent une grande colonne thermique, un grand reservoir d'eau blinde pour les experiences, et une installation de production de radioisotopes munie de tubes de transfert a fonctionnement pneumatique. Trois tubes traversent le coeur central, un autre traverse la colonne thermique. Il existe eniin plusieurs petites gouttieres verticales centrales et une gouttiere verticale a section carree de 15 cm

The main objectives of this study were to describe the current status of the energetic and industrial bio-product markets (biofuels, bio-lubricants, biomaterials, papers, cosmetics, and so on), to identify and analyze the evolution perspectives of these new markets on a long and medium term, to define scenarios of evolution for different sectors (agro-industry, energy, organic chemistry), to identify the most promising new markets, and to select the priority agro-industrial sectors

In order to answer the modeling needs of manufacturers and users of electrical resistance furnaces, Electricite de France (EdF) has been appealed to search for a numerical simulation tool for the modeling of thermal phenomena and to carry out its qualification. The TNG software has been retained according to its modeling characteristics of radiant heat transfers and to the coupling with thermal conduction. After a description of the main characteristics of this code, two examples of application to electrical furnaces are presented. The first example illustrates how it can be possible to accurately reproduce the behaviour of a big industrial furnace. The second example is an illustration of numerical simulation possibilities for the optimization of processes performed with an electric furnace. (J.S.)

The direction of studies and researches (DER) of Electricite de France (EdF) has been involved for several years in a research program on turbulent reactive flows. The objectives of this program concern: the reduction of pollutant emissions from existing fossil-fueled power plants, the study of new production means (fluidized beds), and the promotion of electric power applications in the industry. An important part of this program is devoted to the development and validation of 3-D softwares and to the modeling of physical phenomena. This paper presents some industrial applications (furnaces, boilers, electric arcs) for which radiant heat transfers play an important role and the radiation models used. (J.S.) 8 refs.

This document briefly presents the Cigeo project which is designed for the underground geological storage of the most radioactive wastes. Requirements comprise safety after closure and without any human intervention, and a reversible operation during at least 100 years. The storage principle is briefly described. A brief history of this research project is reported

The French Act of 28 June 2006 identifies reversible disposal in deep geological facilities as the benchmark solution for long-term management of high-level waste (HLW) and for intermediate-level long-lived waste (ILW-LL). The Act tasks ANDRA (national agency for the management of radioactive wastes) with the pursuit of studies and research on the choice of a site and the design of the repository, with a view to examining the licence application in 2015 and, provided that the licence is granted, to make the facility operational by 2025. At the end of 2009, ANDRA submitted to the Government its proposals regarding the site and the design of the Industrial Centre for Geological Disposal, known as CIGEO. With the definition of a possible area for the construction of underground disposal facilities, one of the key stages in the project has been achieved. The choice of a surface site will be validated following the public consultation scheduled for the end of 2012. The project is now on the point of entering the definition stage (preliminary design). CIGEO will be a nuclear facility unlike any other. It will be built and operated for a period of over 100 years. For it to be successful, the project must meet certain requirements related to its integration in the local area, industrial planning, safety and reversibility, while also controlling costs. Reversibility is a very important concept that will be defined by law. It is ANDRA's responsibility to ensure that a reasonable balance is found between these different concerns. (author)

The present report gives an account of results on a research work about 'the stakes of the dialogue around the follow up of nuclear and non nuclear industrial facilities' and on conclusions of a seminar, on the same subject that stood at Ville D' Avray from the 21. to 22. of January 2003. This seminar has gathered different actors (administration, experts, associations, industrial operators) concerned by the dialogue around these installations. The work has been directed by I.R.S.N. and had for object to give the knowledge of the French and International experience in matter of dialogue around nuclear and non nuclear industrial sites. (N.C.)

Cryomagnetic separation adaptation to environment technologies application to industrial liquid effluents. The performance, obtained by superconducting high filed - high gradient magnetic separation, permitted to foresee the magnetic treatment of heavy metals in rinse waters, derived from the surface finishing industry. The paramagnetic ions, precipitated in basic media as hydroxides, present a very hydrated amorphous structure, which masks their subjacent magnetic properties. Coprecipitation of a `magnetic carrier`, jointly with the heavy metals, has been studied: ferric chloride forms in basic media, an hydrated iron oxide. Its structure is of the goethite type, and it stabilizes as hematite. The magnetic susceptibility of the obtained product is still weak and its crystalline structure is not enough affirmative to utilize magnetic filtration with efficiency. Mixture of ferrous sulphate and ferric chloride forms, in a basic media, an hydrated magnetite. Initial ideal ratio between divalent iron and trivalent iron, varies between 0,5 and 1,2. This mixture, coprecipitated with the heavy metals, permits to optimize the magnetic cleaning of the fluids in a high field - high gradient filter. (author)

This work is devoted to the development of a process of detection in real time of metallic pollutants present in industrial atmospheric effluents. The method of measurement is the atomic spectrometry of emission coupled to an ICP torch (Inductively coupled Plasma). The technology of the fluidized beds is used as system of introduction of the metallic particles into the ICP torch, the interest of the principle of detection resting on the stamping from the usual procedure of calibration of the analytical system. The results are presented in two parts. The first relates to the diagnosis of plasmas formed with various mixtures of N{sub 2}/O{sub 2} which one corresponds to pure air, the second presents the setting process of detection in real time starting from the intensities ratios of the spectral lines of the metallic element with those of the plasma-producing element (argon or pure air) The study of the diagnosis of plasmas made up of mixtures N{sub 2}/O{sub 2} relates to the determination of the atomic excitation temperature from the spectral lines of the copper element and the evaluation of the thermal disequilibrium q Te/Th. This last is obtained by considering the mass enthalpy of various mixtures N{sub 2}/O{sub 2}. The existence of a small thermal disequilibrium is highlighted. The study of detection in real time by ICP torch, without calibration of the system, is based on three points: - spectroscopic data processing to determine the values of the intensities ratios of spectral lines; - the insertion of the intensities ratios and the characteristics of plasma (argon or pure air) into a calculation code of plasma composition; - the comparison of the mass flux values of the metallic pollutants, in real time, obtained by experiments with those resulting from the elutriation calculation, term which defines the phenomenon of entrainment of the particles out of the fluidized bed. The results made it possible to show the similarity of the analytical system response between the use of argon plasma and that of air. (author)

In this thesis we will question the perceptions of industrial risks in the occidental world at the beginning of the 21. century. For this purpose we will try to understand how concepts such as sustainable development, precautionary principle, liability, or even zero-risk bias have progressively developed around a thought model based on the scientific rationality. This model is now undermined by its incapacity to fully address the issues it raises and completely avoid the potential risks. However, despite consistent weaknesses, it remains a reference value moulded by past accidents which have led to the making of laws aiming mainly at defining liability and protecting those who are held liable. Thus, public information becomes a requirement for democracy and the protection of this thought model. In this context, the protagonists at stake are security-conscious, economical and political lobbies that constantly redefine the limits of risk acceptance. We come to the realization that our lifestyle and value system remain unchallenged even though undergoing a crisis. The specificity of this research lies into the importance we give to the local approach, dealing with registered Seveso sites and nuclear plants located in Indre et Loire. We have polled five categories of respondents through interviews or questionnaires in order to understand their opinion regarding situations involving technological risks. The result of this survey helps us understand and set the levels of risk acceptance that they define with regard to the industrial risks and show the complexity of a situation involving political stakes, environmental pressures, a profit-driven economy and security constraints, in a vague and complex context. This work gives us a contrasted picture of today's perceptions of risks. (author)

Occurrence of cyanide compounds, mainly [Fe(CN)6]2-/3-, in soils and groundwater is due to industrial waste deposits. The main polluting source are purifier wastes stored on former manufactured gas plant sites. In order to estimate the environmental risks associated with cyanide, a specific analytical procedure combining analyses on the solids (optical and scanning electron microscopies, infrared spectroscopy and x-ray diffraction) with analyses carried out on leachates obtained by aqueous extraction at various pHs was developed. For a given purifier waste, Prussian blue was evidenced as the major cyanide species but other minor compounds were also observed. Some of them are extremely soluble (e.g. potassium ferrocyanide) whereas others are very stable (e.g. potassium and zinc ferrocyanide). The study of other polluted solids allowed to predict the hazards linked with their storage. Measurements realized on percolation columns effluents revealed that cyanide species can migrate as colloidal species as well as under soluble forms, which confirmed some site observations where cyanide pollution was observed in soils located beneath a priori stable wastes. Such a combination of techniques can then be considered as a useful diagnosis and risk assessment tool. However, as the full procedure is rather time-consuming a partial combination of the techniques developed can even be used advantageously in some specific cases. We also developed a quantification and speciation method (spectrometric UV-Visible - mathematical code) to determine cyanide compounds concentrations in water extraction filtrates. Finally, irradiation tests of purifier wastes were carried out. They revealed that the influence of natural light could be considered as negligible in comparison to the other factors affecting cyanide compounds migration.

In the contribution under consideration, the Deutsche Energie-Agentur GmbH (Berlin, Federal Republic of Germany) reports on an energetic modernization of industrial heat supply systems. Possibilities of an enhancement of the energetic efficiency and energy conservation at large combustion engineering plants are described. After an introduction to this theme, the author of this contribution provides an overview of the optimization of heat supply systems, and reports on the following aspects: Optimisation of the heat demand; energy efficient heat generation; heat recovery; energy efficient conversion technology and generation technology; associate partners for more energy efficiency in industry and commerce; best practice examples.

In order to answer the modeling needs of manufacturers and users of electrical resistance furnaces, Electricite de France (EdF) has been appealed to search for a numerical simulation tool for the modeling of thermal phenomena and to carry out its qualification. The TNG software has been retained according to its modeling characteristics of radiant heat transfers and to the coupling with thermal conduction. After a description of the main characteristics of this code, two examples of application to electrical furnaces are presented. The first example illustrates how it can be possible to accurately reproduce the behaviour of a big industrial furnace. The second example is an illustration of numerical simulation possibilities for the optimization of processes performed with an electric furnace. (J.S.)

The direction of studies and researches (DER) of Electricite de France (EdF) has been involved for several years in a research program on turbulent reactive flows. The objectives of this program concern: the reduction of pollutant emissions from existing fossil-fueled power plants, the study of new production means (fluidized beds), and the promotion of electric power applications in the industry. An important part of this program is devoted to the development and validation of 3-D softwares and to the modeling of physical phenomena. This paper presents some industrial applications (furnaces, boilers, electric arcs) for which radiant heat transfers play an important role and the radiation models used. (J.S.) 8 refs.

This final report for the Swiss Federal Office of Energy (SFOE) presents the results of the second phase of the development of a small-scale gas turbine that can use lean gas. The aim of this project phase - the design of a project development unit (PDU) for a micro-turbine with a nominal grid feeding power of 2.4 kWe is discussed. Parallel to this work, peripheral components such as burner and recovery device shall also be designed, produced and tested in the laboratories of the Swiss Center for Appropriate Technology and Social Ecology. The burner is specially designed for the combustion of lean gases. The goals of the following project phase, including the production and field-testing of the unit are discussed. On the basis of this PDU, a prototype lean gas micro-turbine with 9 kWe electrical generation power is to be derived. Project goals, work done and results obtained are reviewed, as is further work to be done.

This document provides information, from 1970 to 2005, on the evolution of the energy intensity (ratio between the primary energy consumption and the gross domestic product in volume) and the actions of energy control for the industrial, residential and transportation sectors. (A.L.B.)

Wheat is the second largest crop cultivated around the world and constitutes a major part of the daily diet in Europe. During the course of improving the baking quality of wheat cultivar, most of the nutritional attributes have been underestimated. It is therefore unfortunate that most of wheat-based food products are mostly produced from refined white flour from which peripheral tissues (germ and envelopes) are removed. However, these tissues, which are eliminated and serve mainly for animal...

Butagaz company has developed a new technical solution to reinforce the safety of its propane storage site of Aumale. This solution is based on the construction of a cylindrical reinforced concrete envelope around the spherical gas tank in order to greatly reduce the risk of boiling liquid expanding vapor explosion (BLEVE). Short paper. (J.S.)

This work deals with the selective catalytic reduction of nitrogen oxides (NO{sub x}), contained in the effluents of industrial plants, by hydrogen or methane. The aim is to replace ammonia, used as reducing agent, in the conventional process. The use of others reducing agents such as hydrogen or methane is interesting for different reasons: practical, economical and ecological. The catalyst has to convert selectively NO into N{sub 2}, in presence of an excess of oxygen, steam and sulfur dioxide. The developed catalyst is constituted by a support such as perovskites, particularly LaCoO{sub 3}, on which are dispersed noble metals (palladium, platinum). The interaction between the noble metal and the support, generated during the activation of the catalyst, allows to minimize the water and sulfur dioxide inhibitor phenomena on the catalytic performances, particularly in the reduction of NO by hydrogen. (O.M.)

Final report concludes the importance of establishing the Research and Development program called 'Gassmaks'. The target of this program is increased value added to the natural gas loop. Strengthened know-how, industrial development and international competition force shall contribute to higher value added to community through industrial refining of natural gas. Gassmaks will by research based foundation exploit Norwegian natural gas resources environmental friendly. Highly prioritised are converting and use of natural gas to plastic raw materials, synthesis gas, synthetical fuel, energy processes, carbon materials, metallurgical processes and nutrients as proteins and fat. (AG). 28 refs., 3 figs., 3 tabs

The main objective of the thesis is the study of a new roof thermal behaviour with the thermal transfer by radiation and convection in order to propose an alternative to the mechanical air-conditioning in the overseas islands, and to determine the way of improving the energy performance. Two fundamental results were obtained. First, the coating has a great importance in relation with the radiative properties. Secondly, air channel geometries can complete insufficient coatings in their cooling purpose. In this thesis, both experimental and modelling analyses were carried out. A solar simulator was used for the experiment to carry out a comparative study of prototypes. A performance indicator was defined; it is the eliminated flux, because the temperature is not sufficient to describe the heat exchange. At the same time, the radiative properties of coatings were measured and a characterisation of paints allows to start a new research of innovative infrared reflective paints. A mono-dimensional model was developed describing thermal exchanges in an air channel component. A sensibility study was carried out which allows to determine the main parameters. The experimental study validated the model in rear flux, more realistic than stagnation. Finally, the feasibility of this new steel component was concluded by a technical-economical study. (author)

Full Text Available Socio-economic factors affecting the use of agro-industrial by-products for cattle fattening in the dry season in Adamawa, Cameroon. Cattle production is the major economic activity in the Adamawa. Feed deficiency that causes about 129 to 187 g weight loss per day is an important constraint during the dry season. A possible alternative to overcome this constraint is the use of agro-industrial by-products. However, the adoption and effective use of these potentials are still to be encouraged. This study, which objective was to find out the socio-economic factors that may affect the usage of these resources, permitted to show, through interviews and surveys, that more than 8,200 t of agro-industrial by-products (maize and wheat bran, soybean seed and maize seed cake useable for cattle feeding were produced annually in Adamawa and that only 16 % of this production were used by livestock farmers. Among the factors affecting the effective use of these by-products, prices were the most determinant. The herds size and the number of sedentary animals had also shown a significant positive effect. On the other hand, livestock farming experience and the farmer’s age had instead shown a strong negative correlation with the use of by-products; same was the level of farmer’s education which showed positive correlation only with respect to cotton seed cake use. The distance from the livestock farming sites to the by-products production/distribution centre and/or difficult access to the sites had strong positive correlations with the by-products’ prices. The problem of supplies as a result of long distance or of difficult access to cattle production sites, the high and unstable prices of agro-industrial by-products and poor awareness of the different types of by-products produced in the Adamawa, were noted as major constraints for their use as cattle feed. Based on the above results, an effective extension system and well organized farmers groupings could improve the level of utilization of agro-industrial by-products for cattle feeding.

This colloquium was jointly organized by the French institute of petroleum (IFP), the French agency of environmental and energy mastery (Ademe) and the geological and mining research office (BRGM). This press kit makes a status of the advances made in CO{sub 2} emissions abatement and geological sequestration: technological advances of CO{sub 2} capture and sequestration, geological reservoir dimensioning with respect to the problem scale, duration of such an interim solution, CO{sub 2} emissions abatement potentialities of geological sequestration, regulatory, economical and financial implications, international stakes of greenhouse gas emissions. This press kit comprises a press release about the IFP-Ademe-BRGM colloquium, a slide presentation about CO{sub 2} abatement and sequestration, and four papers: a joint IFP-Ademe-BRGM press conference, IFP's answers to CO{sub 2} emissions abatement, Ademe's actions in CO{sub 2} abatement and sequestration, and BRGM's experience in CO{sub 2} sequestration and climatic change expertise. (J.S.)

These days are dedicated to the implementation of the radiation protection optimization in the activities of the electronuclear sector, of the industrial sector, the medical sector, the laboratories and the centers of research and the university sector. All the aspects of the practical application of the radiation protection optimization of the workers, the public and the patients will be approached. The oral communications and posters concern the following subjects: foundations of the optimization principle, new statutory context, transmission of ALARA principle, operational dosimetry, conception, operating and maintenance of the installations, the construction sites of dismantling, industrial radiology, radioactive waste management. (N.C.)

An international symposium on the reduction of emissions and geological storage of CO{sub 2} was held in Paris from 15 to 16 September 2005. The event, jointly organized by IFP, ADEME and BRGM, brought together over 400 people from more than 25 countries. It was an opportunity to review the international stakes related to global warming and also to debate ways of reducing CO{sub 2} emissions, taking examples from the energy and transport sectors. The last day was dedicated to technological advances in the capture and geological storage of CO{sub 2} and their regulatory and economic implications. This document gathers the available transparencies and talks presented during the colloquium: Opening address by F. Loos, French Minister-delegate for Industry; Session I - Greenhouse gas emissions: the international stakes. Outlook for global CO{sub 2} emissions. The global and regional scenarios: Alternative scenarios for energy use and CO{sub 2} emissions until 2050 by C. Mandil and J. Podkanski (IEA), The stabilization of CO{sub 2} emissions in the coming 50 years by R. Socolow (Princeton University). Evolution of the international context: the stakes and 'factor 4' issues: Costs of climate impacts and ways towards 'factor 4' by D. Dron (ENS Mines de Paris), CO{sub 2} emissions reduction policy: the situation in the United States by D. Reiner (MIT/Cambridge University), Post-Kyoto scenarios by P. Horrocks (European Commission), Possibilities for R and D in CO{sub 2} capture and storage in the future FP7 program by P. Fernandez Ruiz and P. Dechamps (European Commission). Session II - CO{sub 2} emission reductions in the energy and transport sectors. Reducing CO{sub 2} emissions during the production and conversion of fossil energies (fixed installations): Combined cycles using hydrogen by G. Haupt (Siemens), CO{sub 2} emission reductions in the oil and gas industry by I. Wright (BP). Reducing CO{sub 2} emissions in the transport sector: Sustainable transport systems by P. Wiederkehr (EST International), The prospects for reducing CO{sub 2} emissions in the transport sector (cars and aviation) by A. Morcheoine (ADEME), The contribution of biofuels and alternative fuels to reducing CO{sub 2} emissions in the transport sector by I. Drescher (Volkswagen AG). Session III - Technological progress in the capture and geological storage of CO{sub 2}: European projects on CO{sub 2} capture and storage by P. Dechamps (European Commission, Research Energy Conversion and Transport); Capture of CO{sub 2}: Innovative CO{sub 2} capture concepts by P. Feron (TNO), Capture of CO{sub 2} in pre- and oxy-combustion by A. Feraud, N. Otter (Alstom); Geological storage of CO{sub 2}: Geological storage capacity by N.P. Christensen (GEUS), Feedback from industrial CO{sub 2} storage projects by T. Torp (Statoil), The main avenues of current research by P. Le Thiez (IFP) and I. Czernichowski (BRGM), Long-term industrial experience with underground gas storage by J. Hartman (GDF). Session IV - Regulatory, economic and financial aspects. Legal and regulatory framework for capture and geological storage: UK's perspective on the regulatory framework for CO{sub 2} storage by J. Roberts (DEFRA-UK), Monitoring and reporting of CCS in the European Union Emission Trading Scheme by P. Zakkour (ERM), Social need and public questions and perceptions by G. von Goerne (Greenpeace); Economic and financial impact: The costs of CCS by G. Allinson (CO{sub 2}-CRC), The characteristics of CO{sub 2} markets: players, volumes exchanged, and term and spot transaction prices by L. Segalen (European Carbon Fund), CO{sub 2} management by J.M. Gires (Total), The forthcoming IPCC special report on carbon dioxide capture and storage by B. Metz (IPCC Working Group III). Closing Address by Bernard Brillet, Ministry for Higher Education and Research. (J.S.)

The present thesis deals with industrialization which takes place within press distribution systems in certain countries of European Union., in particular France, Great Britain, Germany, Italy and Spain. It stresses liberalization and regulation of realization operation of titles diffusion according to mechanisms and processes borrowed from industrial logistics and marketing distributor specific to the sector of the current goods distribution. After introduction constraints and specificities ...

POWEO, the leading independent energy operator in France, presents in this document its key financial data and highlights for the first half of 2006: - Half-year revenue amounts to euro 119.4 m, multiplied by 3.4 compared to the same period last year; - The Energy Management activity has achieved a net margin of euro 34.3 m; - EBIT amounts to euro 6.2 m, compared to euro -2.9 m in the first half of 2005; - Net income amounts to euro 8.9 m, compared to euro -2.9 m in the first half of 2005; - Completion of the preliminary steps to the building of a first thermal power plant (CCGT) is close at hand, two other projects launched; - Strengthening of internal structures in view of the residential market opening up; - Outlook for 2006: total sales expected to reach euro 220 m and positive EBITDA; - LNG terminal building project in Le Havre.

Processed baby foods designed for infants (4-12 months) and toddlers (12-36 months) (excluding infant formula, follow-on formula, the so-called growing-up milks, and cereal-based foods for infants), which am referred to as baby foods, are specific products defined by a European regulation (Directive 2006/125/CE). According to this Directive, such foods have a composition adapted to the nutritional needs of children of this age and should comply with specifications related to food safety in te...

This thesis presents an approach to monitoring the controlled discrete events systems. The study is limited to the interruptible faults : intermittent and permanent. To increase the availability of a system, it is crucial to reduce the unnecessary interruptions. For that, the acceptable behavior of these systems is introduced. This behavior presents a tolerance to the intermittent faults. An approach of the construction of monitoring system is presented. We model, firstly, the behavior of the...

The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal d